National Library of Energy BETA

Sample records for liquor spent sulfite

  1. Gasification of black liquor

    DOE Patents [OSTI]

    Kohl, Arthur L. (Woodland Hills, CA)

    1987-07-28

    A concentrated aqueous black liquor containing carbonaceous material and alkali metal sulfur compounds is treated in a gasifier vessel containing a relatively shallow molten salt pool at its bottom to form a combustible gas and a sulfide-rich melt. The gasifier vessel, which is preferably pressurized, has a black liquor drying zone at its upper part, a black liquor solids gasification zone located below the drying zone, and a molten salt sulfur reduction zone which comprises the molten salt pool. A first portion of an oxygen-containing gas is introduced into the gas space in the gasification zone immediatley above the molten salt pool. The remainder of the oxygen-containing gas is introduced into the molten salt pool in an amount sufficient to cause gasification of carbonaceous material entering the pool from the gasification zone but not sufficient to create oxidizing conditions in the pool. The total amount of the oxygen-containing gas introduced both above the pool and into the pool constitutes between 25 and 55% of the amount required for complete combustion of the black liquor feed. A combustible gas is withdrawn from an upper portion of the drying zone, and a melt in which the sulfur content is predominantly in the form of alkali metal sulfide is withdrawn from the molten salt sulfur reduction zone.

  2. Gasification of black liquor

    DOE Patents [OSTI]

    Kohl, A.L.

    1987-07-28

    A concentrated aqueous black liquor containing carbonaceous material and alkali metal sulfur compounds is treated in a gasifier vessel containing a relatively shallow molten salt pool at its bottom to form a combustible gas and a sulfide-rich melt. The gasifier vessel, which is preferably pressurized, has a black liquor drying zone at its upper part, a black liquor solids gasification zone located below the drying zone, and a molten salt sulfur reduction zone which comprises the molten salt pool. A first portion of an oxygen-containing gas is introduced into the gas space in the gasification zone immediately above the molten salt pool. The remainder of the oxygen-containing gas is introduced into the molten salt pool in an amount sufficient to cause gasification of carbonaceous material entering the pool from the gasification zone but not sufficient to create oxidizing conditions in the pool. The total amount of the oxygen-containing gas introduced both above the pool and into the pool constitutes between 25 and 55% of the amount required for complete combustion of the black liquor feed. A combustible gas is withdrawn from an upper portion of the drying zone, and a melt in which the sulfur content is predominantly in the form of alkali metal sulfide is withdrawn from the molten salt sulfur reduction zone. 2 figs.

  3. Causticizing for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    Scott Sinquefeld; James Cantrell; Xiaoyan Zeng; Alan Ball; Jeff Empie

    2009-01-07

    The cost-benefit outlook of black liquor gasification (BLG) could be greatly improved if the smelt causticization step could be achieved in situ during the gasification step. Or, at a minimum, the increase in causticizing load associated with BLG could be mitigated. A number of chemistries have been proven successful during black liquor combustion. In this project, three in situ causticizing processes (titanate, manganate, and borate) were evaluated under conditions suitable for high temperature entrained flow BLG, and low temperature steam reforming of black liquor. The evaluation included both thermodynamic modeling and lab experimentation. Titanate and manganate were tested for complete direct causticizing (to thus eliminate the lime cycle), and borates were evaluated for partial causticizing (to mitigate the load increase associated with BLG). Criteria included high carbonate conversion, corresponding hydroxide recovery upon hydrolysis, non process element (NPE) removal, and economics. Of the six cases (three chemistries at two BLG conditions), only two were found to be industrially viable: titanates for complete causticizing during high temperature BLG, and borates for partial causticizing during high temperature BLG. These two cases were evaluated for integration into a gasification-based recovery island. The Larsen [28] BLG cost-benefit study was used as a reference case for economic forecasting (i.e. a 1500 tpd pulp mill using BLG and upgrading the lime cycle). By comparison, using the titanate direct causticizing process yielded a net present value (NPV) of $25M over the NPV of BLG with conventional lime cycle. Using the existing lime cycle plus borate autocausticizing for extra capacity yielded a NPV of $16M.

  4. REFRACTORY FOR BLACK LIQUOR GASIFIERS

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Jun Wei

    2005-03-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. This report covers Task 1.3, Simulative corrosion of candidate materials developed by refractory producers and in the laboratory based on the results of Task 1.1 and Task 1.2. Refractories provided by in-kind sponsors were tested by cup testing, density/porosity determinations, chemical analysis and microscopy. The best performing materials in the cup testing were fused cast materials. However, 2 castables appear to outperforming any of the previously tested materials and may perform better than the fused cast materials in operation. The basis of the high performance of these materials is the low open porosity and permeability to black liquor smelt.

  5. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    William L. Headrick Jr; Alireza Rezaie; Xiaoting Liang; Musa Karakus; Jun Wei

    2005-12-01

    The University of Missouri-Rolla identified materials that permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project was to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study attempted to define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials were selected or developed that reacted with the gasifier environment to form protective surfaces in-situ; and were functionally-graded to give the best combination of thermal, mechanical and physical properties and chemical stability; and are relatively inexpensive, reliable repair materials. Material development was divided into 2 tasks: Task 1 was development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesium aluminate and barium aluminate for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO2 and SiC. Task 2 was finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  6. REFRACTORY FOR BLACK LIQUOR GASIFIERS

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Xiaoting Liang; Jun Wei

    2005-04-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesium aluminate and barium aluminate for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO{sub 2} and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  7. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Xiaoting Liang

    2005-10-01

    The University of Missouri-Rolla identified materials that permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project was to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study attempted to define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials were selected/developed that either react with the gasifier environment to form protective surfaces in-situ; and were functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development were divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesium aluminate and barium aluminate for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO{sub 2} and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  8. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    William L. Headrick Jr; Alireza Rezaie

    2003-08-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the materials problems encountered during the operation of low-pressure high-temperature (LFHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesia aluminate and baria aluminate spinels for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  9. REFRACTORY FOR BLACK LIQUOR GASIFIERS

    SciTech Connect (OSTI)

    William L. Headrick Jr.; Alireza Rezaie

    2003-12-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the materials problems encountered during the operation of low-pressure high-temperature (LFHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesia aluminate and baria aluminate spinels for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  10. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    William L. Headrick Jr; Alireza Rezaie

    2003-12-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the materials problems encountered during the operation of low-pressure high-temperature (LFHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesia aluminate and baria aluminate spinels for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  11. REFRACTORY FOR BLACK LIQUOR GASIFIERS

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Xiaoting Liang

    2005-07-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesium aluminate and barium aluminate for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO{sub 2} and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  12. REFRACTORY FOR BLACK LIQUOR GASIFIERS

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Xiaoting Liang; Jun Wei

    2005-01-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesium aluminate and barium aluminate for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO{sub 2} and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  13. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    Robert E. Moore; William L. Headrick; Alireza Rezaie

    2003-03-31

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the materials problems encountered during the operation of low-pressure high-temperature (LFHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. Material development will be divided into 2 tasks: Task 1, Development and property determinations of improved and existing refractory systems for black liquor containment. Refractory systems of interest include magnesia aluminate and baria aluminate spinels for binder materials, both dry and hydratable, and materials with high alumina contents, 85-95 wt%, aluminum oxide, 5.0-15.0 wt%, and BaO, SrO, CaO, ZrO and SiC. Task 2, Finite element analysis of heat flow and thermal stress/strain in the refractory lining and steel shell of existing and proposed vessel designs. Stress and strain due to thermal and chemical expansion has been observed to be detrimental to the lifespan of existing black liquor gasifiers. The thermal and chemical strain as well as corrosion rates must be accounted for in order to predict the lifetime of the gasifier containment materials.

  14. Refractory for Black Liquor Gasifiers

    SciTech Connect (OSTI)

    William L. Headrick Jr; Musa Karakus; Xiaoting Laing

    2005-10-01

    The University of Missouri-Rolla will identify materials that will permit the safe, reliable and economical operation of combined cycle gasifiers by the pulp and paper industry. The primary emphasis of this project will be to resolve the material problems encountered during the operation of low-pressure high-temperature (LPHT) and low-pressure low-temperature (LPLT) gasifiers while simultaneously understanding the materials barriers to the successful demonstration of high-pressure high-temperature (HPHT) black liquor gasifiers. This study will define the chemical, thermal and physical conditions in current and proposed gasifier designs and then modify existing materials and develop new materials to successfully meet the formidable material challenges. Resolving the material challenges of black liquor gasification combined cycle technology will provide energy, environmental, and economic benefits that include higher thermal efficiencies, up to three times greater electrical output per unit of fuel, and lower emissions. In the near term, adoption of this technology will allow the pulp and paper industry greater capital effectiveness and flexibility, as gasifiers are added to increase mill capacity. In the long term, combined-cycle gasification will lessen the industry's environmental impact while increasing its potential for energy production, allowing the production of all the mill's heat and power needs along with surplus electricity being returned to the grid. An added benefit will be the potential elimination of the possibility of smelt-water explosions, which constitute an important safety concern wherever conventional Tomlinson recovery boilers are operated. Developing cost-effective materials with improved performance in gasifier environments may be the best answer to the material challenges presented by black liquor gasification. Refractory materials may be selected/developed that either react with the gasifier environment to form protective surfaces in-situ; are functionally-graded to give the best combination of thermal, mechanical, and physical properties and chemical stability; or are relatively inexpensive, reliable repair materials. This report covers Task 1.4, Industrial Trial of candidate materials developed by refractory producers and in the laboratory based on the results of Task 1.1, 1.2 and 1.3. Refractories provided by in-kind sponsors to industrial installations tested by cup testing, density/porosity determinations, chemical analysis and microscopy. None of the materials produced in this program have been tried in high temperature gasifiers, but the mortar developed Morcocoat SP-P is outperforming other mortars tested at ORNL. MORCO PhosGun M-90-O has shown in laboratory testing to be an acceptable candidate for hot and cold repairs of existing high temperature gasifiers. It may prove to be an acceptable lining material.

  15. Value recovery from spent alumina-base catalyst

    DOE Patents [OSTI]

    Hyatt, David E. (Northglenn, CO)

    1987-01-01

    A process for the recovery of aluminum and at least one other metal selected from the group consisting of molybdenum, nickel and cobalt from a spent hydrogenation catalyst comprising (1) adding about 1 to 3 parts H.sub.2 SO.sub.4 to each part of spent catalyst in a reaction zone of about 20.degree. to 200.degree. C. under sulfide gas pressure between about 1 and about 35 atmospheres, (2) separating the resultant Al.sub.2 (SO.sub.4).sub.3 solution from the sulfide precipitate in the mixture, (3) oxidizing the remaining sulfide precipitate as an aqueous slurry at about 20.degree. to 200.degree. C. in an oxygen-containing atmosphere at a pressure between about 1 and about 35 atmospheres, (4) separating the slurry to obtain solid molybdic acid and a sulfate liquor containing said at least one metal, and (5) recovering said at least one metal from the sulfate liquor in marketable form.

  16. Proceedings of the black liquor research program review fourth meeting held July 28--30, 1987

    SciTech Connect (OSTI)

    Emerson, D. B.; Whitworth, B. A.

    1987-10-01

    Research programs, presented at the black liquor review meeting are described. Research topics include the following: Cooperative Program in Kraft Recovery; Black Liquor Physical Properties; Viscosity of Strong Black Liquor; Ultrafiltration of Kraft Black Liquor; Molecular Weight Distribution of Kraft Lignin; Black Liquor Droplet Formation Project; Fundamental Studies of Black Liquor Combustion; Black Liquor Combustion Sensors; Flash X-ray Imagining of Black Liquor Sprays; Laser Induced Fluorescence For Process Control In The Pulp and Paper Industry; Recovery Boiler Optimization; Black Liquor Gasification and Use of the Products in Combined-Cycle Cogeneration; Black Liquor Steam Plasma Automization; The B and W Pyrosonic 2000R System; Monsteras Boiler Control System; and Cooperative Program Project Reviews. Individual projects are processed separately for the data bases.

  17. Value-Added Products from FGD Sulfite-Rich Scrubber Materials

    SciTech Connect (OSTI)

    Vivak Malhotra

    2010-01-31

    According to the American Coal Ash Association, about 29.25 million tons of flue gas desulfurization (FGD) byproducts were produced in the USA in 2003. Out of 29.25 million tons, 17.35 million tons were sulfite-rich scrubber materials. At present, unlike its cousin FGD gypsum, the prospect for effective utilization of sulfite-rich scrubber materials is not bright. In fact, almost 16.9 million tons are leftover every year. In our pursuit to mitigate the liability of sulfite-rich FGD scrubber materials' disposal, we are attempting to develop value-added products that can commercially compete. More specifically, for this Innovative Concept Phase I project, we have the following objectives: to characterize the sulfite-rich scrubber material for toxic metals; to optimize the co-blending and processing of scrubber material and natural byproducts; to formulate and develop structural composites from sulfite-rich scrubber material; and to evaluate the composites' mechanical properties and compare them with current products on the market. After successfully demonstrating the viability of our research, a more comprehensive approach will be proposed to take these value-added materials to fruition.

  18. Drum drying of black liquor using superheated steam impinging jets

    SciTech Connect (OSTI)

    Shiravi, A.H.; Mujumdar, A.S.; Kubes, G.J. [McGill Univ., Montreal, Quebec (Canada)

    1997-05-01

    A novel drum dryer for black liquor utilizing multiple impinging jets of superheated steam was designed and built to evaluate the performance characteristics and effects of various operating parameters thereon. Appropriate ranges of parameters such as steam jet temperature and velocity were examined experimentally to quantify the optimal operating conditions for the formation of black liquor film on the drum surface as well as the drying kinetics.

  19. Applications of pulsed EPR spectroscopy to structural studies of sulfite oxidizing enzymes

    SciTech Connect (OSTI)

    Klein, Eric L.; Astashkin, Andrei V.; Raitsimring, Arnold; Enemark, John H.

    2013-01-01

    Sulfite oxidizing enzymes (SOEs), including sulfite oxidase (SO) and bacterial sulfite dehydrogenase (SDH), catalyze the oxidation of sulfite (SO32?) to sulfate (SO42?). The active sites of SO and SDH are nearly identical, each having a 5-coordinate, pseudo-square-pyramidal Mo with an axial oxo ligand and three equatorial sulfur donor atoms. One sulfur is from a conserved Cys residue and two are from a pyranopterindithiolene (molybdopterin, MPT) cofactor. The identity of the remaining equatorial ligand, which is solvent-exposed, varies during the catalytic cycle. Numerous in vitro studies, particularly those involving electron paramagnetic resonance (EPR) spectroscopy of the Mo(V) states of SOEs, have shown that the identity and orientation of this exchangeable equatorial ligand depends on the buffer pH, the presence and concentration of certain anions in the buffer, as well as specific point mutations in the protein. Until very recently, however, EPR has not been a practical technique for directly probing specific structures in which the solvent-exposed, exchangeable ligand is an O, OH?, H2O, SO32?, or SO42? group, because the primary O and S isotopes (16O and 32S) are magnetically silent (I = 0). This review focuses on the recent advances in the use of isotopic labeling, variable-frequency high resolution pulsed EPR spectroscopy, synthetic model compounds, and DFT calculations to elucidate the roles of various anions, point mutations, and steric factors in the formation, stabilization, and transformation of SOE active site structures.

  20. Molecular Basis for Enzymatic Sulfite Oxidation -- HOW THREE CONSERVED ACTIVE SITE RESIDUES SHAPE ENZYME ACTIVITY

    SciTech Connect (OSTI)

    Bailey, Susan; Rapson, Trevor; Johnson-Winters, Kayunta; Astashkin, Andrei; Enemark, John; Kappler, Ulrike

    2008-11-10

    Sulfite dehydrogenases (SDHs) catalyze the oxidation and detoxification of sulfite to sulfate, a reaction critical to all forms of life. Sulfite-oxidizing enzymes contain three conserved active site amino acids (Arg-55, His-57, and Tyr-236) that are crucial for catalytic competency. Here we have studied the kinetic and structural effects of two novel and one previously reported substitution (R55M, H57A, Y236F) in these residues on SDH catalysis. Both Arg-55 and His-57 were found to have key roles in substrate binding. An R55M substitution increased Km(sulfite)(app) by 2-3 orders of magnitude, whereas His-57 was required for maintaining a high substrate affinity at low pH when the imidazole ring is fully protonated. This effect may be mediated by interactions of His-57 with Arg-55 that stabilize the position of the Arg-55 side chain or, alternatively, may reflect changes in the protonation state of sulfite. Unlike what is seen for SDHWT and SDHY236F, the catalytic turnover rates of SDHR55M and SDHH57A are relatively insensitive to pH (~;;60 and 200 s-1, respectively). On the structural level, striking kinetic effects appeared to correlate with disorder (in SDHH57A and SDHY236F) or absence of Arg-55 (SDHR55M), suggesting that Arg-55 and the hydrogen bonding interactions it engages in are crucial for substrate binding and catalysis. The structure of SDHR55M has sulfate bound at the active site, a fact that coincides with a significant increase in the inhibitory effect of sulfate in SDHR55M. Thus, Arg-55 also appears to be involved in enabling discrimination between the substrate and product in SDH.

  1. Black liquor gasification phase 2D final report

    SciTech Connect (OSTI)

    Kohl, A.L.; Stewart, A.E.

    1988-06-01

    This report covers work conducted by Rockwell International under Amendment 5 to Subcontract STR/DOE-12 of Cooperative Agreement DE-AC-05-80CS40341 between St. Regis Corporation (now Champion International) and the Department of Energy (DOE). The work has been designated Phase 2D of the overall program to differentiate it from prior work under the same subcontract. The overall program is aimed at demonstrating the feasibility of and providing design data for the Rockwell process for gasifying Kraft black liquor. In this process, concentrated black liquor is converted into low-Btu fuel gas and reduced melt by reaction with air in a specially designed gasification reactor.

  2. Advancement of High Temperature Black Liquor Gasification Technology

    SciTech Connect (OSTI)

    Craig Brown; Ingvar Landalv; Ragnar Stare; Jerry Yuan; Nikolai DeMartini; Nasser Ashgriz

    2008-03-31

    Weyerhaeuser operates the world's only commercial high-temperature black liquor gasifier at its pulp mill in New Bern, NC. The unit was started-up in December 1996 and currently processes about 15% of the mill's black liquor. Weyerhaeuser, Chemrec AB (the gasifier technology developer), and the U.S. Department of Energy recognized that the long-term, continuous operation of the New Bern gasifier offered a unique opportunity to advance the state of high temperature black liquor gasification toward the commercial-scale pressurized O2-blown gasification technology needed as a foundation for the Forest Products Bio-Refinery of the future. Weyerhaeuser along with its subcontracting partners submitted a proposal in response to the 2004 joint USDOE and USDA solicitation - 'Biomass Research and Development Initiative'. The Weyerhaeuser project 'Advancement of High Temperature Black Liquor Gasification' was awarded USDOE Cooperative Agreement DE-FC26-04NT42259 in November 2004. The overall goal of the DOE sponsored project was to utilize the Chemrec{trademark} black liquor gasification facility at New Bern as a test bed for advancing the development status of molten phase black liquor gasification. In particular, project tasks were directed at improvements to process performance and reliability. The effort featured the development and validation of advanced CFD modeling tools and the application of these tools to direct burner technology modifications. The project also focused on gaining a fundamental understanding and developing practical solutions to address condensate and green liquor scaling issues, and process integration issues related to gasifier dregs and product gas scrubbing. The Project was conducted in two phases with a review point between the phases. Weyerhaeuser pulled together a team of collaborators to undertake these tasks. Chemrec AB, the technology supplier, was intimately involved in most tasks, and focused primarily on the design, specification and procurement of facility upgrades. Chemrec AB is also operating a pressurized, O2-blown gasifier pilot facility in Piteaa, Sweden. There was an exchange of knowledge with the pressurized projects including utilization of the experimental results from facilities in Piteaa, Sweden. Resources at the Georgia Tech Research Corporation (GTRC, a.k.a., the Institute of Paper Science and Technology) were employed primarily to conduct the fundamental investigations on scaling and plugging mechanisms and characterization of green liquor dregs. The project also tapped GTRC expertise in the development of the critical underlying black liquor gasification rate subroutines employed in the CFD code. The actual CFD code development and application was undertaken by Process Simulation, Ltd (PSL) and Simulent, Ltd. PSL focused on the overall integrated gasifier CFD code, while Simulent focused on modeling the black liquor nozzle and description of the black liquor spray. For nozzle development and testing Chemrec collaborated with ETC (Energy Technology Centre) in Piteae utilizing their test facility for nozzle spray investigation. GTI (Gas Technology Institute), Des Plains, IL supported the team with advanced gas analysis equipment during the gasifier test period in June 2005.

  3. Spent Nuclear Fuel

    Gasoline and Diesel Fuel Update (EIA)

    Nuclear & Uranium Glossary › FAQS › Overview Data Status of U.S. Nuclear Outages (interactive) Summary Uranium & nuclear fuel Nuclear power plants Spent nuclear fuel International All nuclear data reports Analysis & Projections Major Topics Most popular Nuclear plants and reactors Projections Recurring Uranium All reports Browse by Tag Alphabetical Frequency Tag Cloud Previous releases 2002 1998 Spent Nuclear Fuel Release date: December 7, 2015 Next release date: Late 2018 Spent

  4. A comprehensive program to develop correlations for physical properties of kraft black liquor. Final report

    SciTech Connect (OSTI)

    Fricke, A.L.; Zaman, A.A.

    1998-05-01

    The overall objective of the program was to develop correlations to predict physical properties within requirements of engineering precision from a knowledge of pulping conditions and of kraft black liquor composition, if possible. These correlations were to include those relating thermodynamic properties to pulping conditions and liquor composition. The basic premise upon which the research was based is the premise that black liquor behaves as a polymer solution. This premise has proven to be true, and has been used successfully in developing data reduction methods and in interpreting results. A three phase effort involving pulping, analysis of liquor composition, and measurement of liquor properties was conducted.

  5. TEPP- Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel.  This exercise manual is one in...

  6. Redox states of Desulfovibrio vulgaris DsrC, a key protein in dissimilatory sulfite reduction

    SciTech Connect (OSTI)

    Venceslau, Sofia S.; Cort, John R.; Baker, Erin Shammel; Chu, Rosalie K.; Robinson, Errol W.; Dahl, Christiane; Saraiva, Ligia M.; Pereira, Ines Ac

    2013-11-29

    Dissimilatory reduction of sulfite is carried out by the siroheme enzyme DsrAB, with the involvement of the protein DsrC having two conserved cysteine residues. Here, we report a study of the distribution of DsrC in cell extracts, a cysteine-labelling gel-shift assay to monitor its redox state and behaviour, and procedures to produce the different redox forms. We show that, in the model sulfate reducer Desulfovibrio vulgaris, the majority of DsrC is not associated with DsrAB and is thus free to interact with other proteins. In addition, we successfully produced DsrC with an intramolecular disulfide bond (oxidized state) by treatment with arginine.

  7. Spent fuel storage alternatives

    SciTech Connect (OSTI)

    O'Connell, R.H.; Bowidowicz, M.A.

    1983-01-01

    This paper compares a small onsite wet storage pool to a dry cask storage facility in order to determine what type of spent fuel storage alternatives would best serve the utilities in consideration of the Nuclear Waste Policy Act of 1982. The Act allows the DOE to provide a total of 1900 metric tons (MT) of additional spent fuel storage capacity to utilities that cannot reasonably provide such capacity for themselves. Topics considered include the implementation of the Act (DOE away-from reactor storage), the Act's impact on storage needs, and an economic evaluation. The Waste Act mandates schedules for the determination of several sites, the licensing and construction of a high-level waste repository, and the study of a monitored retrievable storage facility. It is determined that a small wet pool storage facility offers a conservative and cost-effective approach for many stations, in comparison to dry cask storage.

  8. Use of sulfide-containing liquors for removing mercury from flue gases

    DOE Patents [OSTI]

    Nolan, Paul S.; Downs, William; Bailey, Ralph T.; Vecci, Stanley J.

    2006-05-02

    A method and apparatus for reducing and removing mercury in industrial gases, such as a flue gas, produced by the combustion of fossil fuels, such as coal, adds sulfide ions to the flue gas as it passes through a scrubber. Ideally, the source of these sulfide ions may include at least one of: sulfidic waste water, kraft caustic liquor, kraft carbonate liquor, potassium sulfide, sodium sulfide, and thioacetamide. The sulfide ion source is introduced into the scrubbing liquor as an aqueous sulfide species. The scrubber may be either a wet or dry scrubber for flue gas desulfurization systems.

  9. Investigation of Fuel Chemistry and Bed Performance in a Fluidized Bed Black Liquor Steam Reformer

    SciTech Connect (OSTI)

    Kevin Whitty

    2007-06-30

    University of Utah's project entitled 'Investigation of Fuel Chemistry and Bed Performance in a Fluidized Bed Black Liquor Steam Reformer' (DOE Cooperative Agreement DE-FC26-02NT41490) was developed in response to a solicitation released by the U.S. Department of Energy in December 2001, requesting proposals for projects targeted towards black liquor/biomass gasification technology support research and development. Specifically, the solicitation was seeking projects that would provide technical support for Department of Energy supported black liquor and biomass gasification demonstration projects under development at the time.

  10. Method for improving separation of carbohydrates from wood pulping and wood or biomass hydrolysis liquors

    DOE Patents [OSTI]

    Griffith, William Louis; Compere, Alicia Lucille; Leitten, Jr., Carl Frederick

    2010-04-20

    A method for separating carbohydrates from pulping liquors includes the steps of providing a wood pulping or wood or biomass hydrolysis pulping liquor having lignin therein, and mixing the liquor with an acid or a gas which forms an acid upon contact with water to initiate precipitation of carbohydrate to begin formation of a precipitate. During precipitation, at least one long chain carboxylated carbohydrate and at least one cationic polymer, such as a polyamine or polyimine are added, wherein the precipitate aggregates into larger precipitate structures. Carbohydrate gel precipitates are then selectively removed from the larger precipitate structures. The method process yields both a carbohydrate precipitate and a high purity lignin.

  11. Use of sulfide-containing liquors for removing mercury from flue gases

    DOE Patents [OSTI]

    Nolan, Paul S. (North Canton, OH); Downs, William (Alliance, OH); Bailey, Ralph T. (Uniontown, OH); Vecci, Stanley J. (Alliance, OH)

    2003-01-01

    A method and apparatus for reducing and removing mercury in industrial gases, such as a flue gas, produced by the combustion of fossil fuels, such as coal, adds sulfide ions to the flue gas as it passes through a scrubber. Ideally, the source of these sulfide ions may include at least one of: sulfidic waste water, kraft caustic liquor, kraft carbonate liquor, potassium sulfide, sodium sulfide, and thioacetamide. The sulfide ion source is introduced into the scrubbing liquor as an aqueous sulfide species. The scrubber may be either a wet or dry scrubber for flue gas desulfurization systems.

  12. Simultaneous and rapid determination of multiple component concentrations in a Kraft liquor process stream

    DOE Patents [OSTI]

    Li, Jian (Marietta, GA); Chai, Xin Sheng (Atlanta, GA); Zhu, Junyoung (Marietta, GA)

    2008-06-24

    The present invention is a rapid method of determining the concentration of the major components in a chemical stream. The present invention is also a simple, low cost, device of determining the in-situ concentration of the major components in a chemical stream. In particular, the present invention provides a useful method for simultaneously determining the concentrations of sodium hydroxide, sodium sulfide and sodium carbonate in aqueous kraft pulping liquors through use of an attenuated total reflectance (ATR) tunnel flow cell or optical probe capable of producing a ultraviolet absorbency spectrum over a wavelength of 190 to 300 nm. In addition, the present invention eliminates the need for manual sampling and dilution previously required to generate analyzable samples. The inventive method can be used in Kraft pulping operations to control white liquor causticizing efficiency, sulfate reduction efficiency in green liquor, oxidation efficiency for oxidized white liquor and the active and effective alkali charge to kraft pulping operations.

  13. Recovery of sugars from ionic liquid biomass liquor by solvent extraction

    Office of Scientific and Technical Information (OSTI)

    (Patent) | SciTech Connect Patent: Recovery of sugars from ionic liquid biomass liquor by solvent extraction Citation Details In-Document Search Title: Recovery of sugars from ionic liquid biomass liquor by solvent extraction The present invention provides for a composition comprising a solution comprising (a) an ionic liquid (IL) or ionic liquid-aqueous (ILA) phase and (b) an organic phase, wherein the solution comprises a sugar and a boronic acid. The present invention also provides for a

  14. Reaction of hydrogen sulfide with oxygen in the presence of sulfite

    SciTech Connect (OSTI)

    Weres, O.; Tsao, L.

    1983-01-14

    Commonly, abatement of hydrogen sulfide emission from a geothermal powerplant requires that hydrogen sulfide dissolved in the cooling water be eliminated by chemical reaction. Oxidation by atmospheric oxygen is the preferred reaction, but requires a suitable catalyst. Nickel is the most potent and thereby cheapest catalyst for this purpose. One mg/L nickel in the cooling water would allow 99% removal of hydrogen sulfide to be attained. A major drawback of catalytic air oxidation is that colloidal sulfur is a major reaction product; this causes rapid sludge accumulation and deposition of sulfur scale. We studied the kinetics and product distribution of the reaction of hydrogen sulfide with oxygen, catalyzed by nickel. Adding sodium sulfite to the solution completely suppresses formation of colloidal sulfur by converting it to thiosulfate. The oxidation reaction is an autocatalytic, free radical chain reaction. A rate expression for this reaction and a detailed reaction mechanism were developed. Nickel catalyzes the chain initiation step, and polysulfidoradical ions propagate the chains. Several complexes of iron and cobalt were also studied. Iron citrate and iron N-hydroxyEDTA are the most effective iron based catalysts. Uncomplexed cobalt is as effective as nickel, but forms a precipitate of cobalt oxysulfide and is too expensive for practical use. 33 figures, 9 tables.

  15. Reaction of Hydrogen Sulfide with Oxygen in the Presence ofSulfite

    SciTech Connect (OSTI)

    Weres, Oleh; Tsao, Leon

    1983-01-01

    Commonly, abatement of hydrogen sulfide emissions from a geothermal powerplant requires that hydrogen sulfide dissolved in the cooling water be eliminated by chemical reaction. Oxidation by atmospheric oxygen is the preferred reaction, but requires a suitable catalyst. Nickel is the most potent and thereby cheapest catalyst for this purpose. One Mg/L nickel in the cooling water would allow 99% removal of hydrogen sulfide to be attained. A major drawback of catalytic air oxidation is that colloidal sulfur is a major reaction product; this causes rapid sludge accumulation and deposition of sulfur scale. The authors studied the kinetics and product distribution of the reaction of hydrogen sulfide with oxygen, catalyzed by nickel. Adding sodium sulfite to the solution completely suppresses formation of colloidal sulfur by converting it to thiosulfate. The oxidation reaction is an autocatalytic, free radical chain reaction. A rate expression for this reaction and a detailed reaction mechanism were developed. Nickel catalyzes the chain initiation step, and polysulfidoradical ions propagate the chains. Several complexes of iron and cobalt were also studied. Iron citrate and iron N-hydroxyEDT are the most effective iron based catalysts. Uncomplexed cobalt is as effective as nickel, but forms a precipitate of cobalt oxysulfide and is too expensive for practical use.

  16. Investigation of Pressurized Entrained-Flow Kraft Black Liquor Gasification in an Industrially Relevant Environment

    SciTech Connect (OSTI)

    Kevin Whitty

    2008-06-30

    The University of Utah's project 'Investigation of Pressurized Entrained-Flow Kraft Black Liquor Gasification in an Industrially Relevant Environment' (U.S. DOE Cooperative Agreement DE-FC26-04NT42261) was a response to U.S. DOE/NETL solicitation DE-PS36-04GO94002, 'Biomass Research and Development Initiative' Topical Area 4-Kraft Black Liquor Gasification. The project began September 30, 2004. The objective of the project was to improve the understanding of black liquor conversion in high pressure, high temperature reactors that gasify liquor through partial oxidation with either air or oxygen. The physical and chemical characteristics of both the gas and condensed phase were to be studied over the entire range of liquor conversion, and the rates and mechanisms of processes responsible for converting the liquor to its final smelt and syngas products were to be investigated. This would be accomplished by combining fundamental, lab-scale experiments with measurements taken using a new semi-pilot scale pressurized entrained-flow gasifier. As a result of insufficient availability of funds and changes in priority within the Office of Biomass Programs of the U.S. Department of Energy, the research program was terminated in its second year. In total, only half of the budgeted funding was made available for the program, and most of this was used during the first year for construction of the experimental systems to be used in the program. This had a severe impact on the program. As a consequence, most of the planned research was unable to be performed. Only studies that relied on computational modeling or existing experimental facilities started early enough to deliver useful results by the time to program was terminated Over the course of the program, small scale (approx. 1 ton/day) entrained-flow gasifier was designed and installed at the University of Utah's off-campus Industrial Combustion and Gasification Research Facility. The system is designed to operate at pressures as high as 32 atmospheres, and at temperatures as high as 1500 C (2730 F). Total black liquor processing capacity under pressurized, oxygen-blown conditions should be in excess of 1 ton black liquor solids per day. Many sampling ports along the conversion section of the system will allow detailed analysis of the environment in the gasifier under industrially representative conditions. Construction was mostly completed before the program was terminated, but resources were insufficient to operate the system. A system for characterizing black liquor sprays in hot environments was designed and constructed. Silhouettes of black liquor sprays formed by injection of black liquor through a twin fluid (liquor and atomizing air) nozzle were videoed with a high-speed camera, and the resulting images were analyzed to identify overall characteristics of the spray and droplet formation mechanisms. The efficiency of liquor atomization was better when the liquor was injected through the center channel of the nozzle, with atomizing air being introduced in the annulus around the center channel, than when the liquor and air feed channels were reversed. Atomizing efficiency and spray angle increased with atomizing air pressure up to a point, beyond which additional atomizing air pressure had little effect. Analysis of the spray patterns indicates that two classifications of droplets are present, a finely dispersed 'mist' of very small droplets and much larger ligaments of liquor that form at the injector tip and form one or more relatively large droplets. This ligament and subsequent large droplet formation suggests that it will be challenging to obtain a narrow distribution of droplet sizes when using an injector of this design. A model for simulating liquor spray and droplet formation was developed by Simulent, Inc. of Toronto. The model was able to predict performance when spraying water that closely matched the vendor specifications. Simulation of liquor spray indicates that droplets on the order 200-300 microns can be expected, and that higher liquor flow will result in be

  17. Active Interrogation for Spent Fuel

    SciTech Connect (OSTI)

    Swinhoe, Martyn Thomas; Dougan, Arden

    2015-11-05

    The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

  18. Generator-Absorber heat exchange transfer apparatus and method using an intermediate liquor

    DOE Patents [OSTI]

    Phillips, Benjamin A. (Benton Harbor, MI); Zawacki, Thomas S. (St. Joseph, MI)

    1996-11-05

    Numerous embodiments and related methods for generator-absorber heat exchange (GAX) are disclosed, particularly for absorption heat pump systems. Such embodiments and related methods use the working solution of the absorption system for the heat transfer medium where the working solution has an intermediate liquor concentration.

  19. Novel Pulping Technology: Directed Green Liquor Utilization (D-GLU) Pulping

    SciTech Connect (OSTI)

    Lucian A. Lucia

    2005-11-15

    The general objectives of this new project are the same as those described in the original proposal. Conventional kraft pulping technologies will be modified for significant improvements in pulp production, such as strength, bleachability, and yield by using green liquor, a naturally high, kraft mill-derived sulfidity source. Although split white liquor sulfidity and other high sulfidity procedures have the promise of addressing several of the latter important economic needs of pulp mills, they require considerable engineering/capital retrofits, redesigned production methods, and thus add to overall mill expenditures. Green liquor use, however, possesses the required high sulfidity to obtain in general the benefits attributable to higher sulfidity cooking, without the required capital constraints for implementation. Before introduction of green liquor in our industrial operations, a stronger understanding of its fundamental chemical interaction with the lignin and carbohydrates in US hardwood and softwoods must be obtained. In addition, its effect on bleachability, enhancement of pulp properties, and influence on the overall energy and recovery of the mill requires further exploration before the process witnesses widespread mill use in North America. Thus, proof of principle will be accomplished in this work and the consequent effect of green liquor and other high sulfide sources on the pulping and bleaching operations will be explored for US kraft mills. The first year of this project will generate the pertinent information to validate its ability for implementation in US pulping operations, whereas year two will continue this work while proceeding to analyze pulp bleachability and final pulp/paper properties and develop a general economic and feasibility analysis for its eventual implementation in North America.

  20. Spent-fuel-storage alternatives

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  1. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

  2. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); Von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Brummond, William A. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  3. Regeneration of FGD waste liquors: Production of ammonium and potassium sulfate mixed fertilizer. Quarterly technical report, July 1993--September 1993

    SciTech Connect (OSTI)

    Randolph, A.D.; Kwon, T.M.

    1993-12-01

    Regeneration of the Fe{sup II}-EDTA scrubbing liquors for simultaneous removal of SO{sub 2} and NO{sub x} in flue gas involves removing the nitrogen-sulfur (N-S) compounds accumulated in the liquor. In this paper, the authors investigated a simple regeneration process which selectively precipitates the N-S compounds as potassium salts and then hydrolyzes them to yield ammonium/potassium sulfate as a marketable byproduct. They believe this is the first report on precipitation and hydrolysis characteristics of the N-S compounds in actual waste scrubbing liquor. Precipitation of the N-S compounds was achieved by adding K{sub 2}SO{sub 4} to the scrubbing liquor. Effects of the amount of added K{sub 2}SO{sub 4} on the amount of removed N-S compounds, precipitated crystals, and the potassium left over in the scrubbing liquor were studied. Hydrolysis of the precipitated N-S compounds to ammonium sulfate was performed in a sulfuric acid environment. Effects of acidity, concentration of N-S compounds, and temperature on the hydrolysis are discussed. Analysis of the observed hydrolysis pattern showed that the reaction proceeded following first order kinetics in terms of N-S compound concentration.

  4. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect (OSTI)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  5. Black liquor combustion validated recovery boiler modeling: Final year report. Volume 4 (Appendix IV)

    SciTech Connect (OSTI)

    Grace, T.M.; Frederick, W.J.; Salcudean, M.; Wessel, R.A.

    1998-08-01

    This project was initiated in October 1990, with the objective of developing and validating a new computer model of a recovery boiler furnace using a computational fluid dynamics (CFD) code specifically tailored to the requirements for solving recovery boiler flows, and using improved submodels for black liquor combustion based on continued laboratory fundamental studies. The key tasks to be accomplished were as follows: (1) Complete the development of enhanced furnace models that have the capability to accurately predict carryover, emissions behavior, dust concentrations, gas temperatures, and wall heat fluxes. (2) Validate the enhanced furnace models, so that users can have confidence in the predicted results. (3) Obtain fundamental information on aerosol formation, deposition, and hardening so as to develop the knowledge base needed to relate furnace model outputs to plugging and fouling in the convective sections of the boiler. (4) Facilitate the transfer of codes, black liquid submodels, and fundamental knowledge to the US kraft pulp industry. Volume 4 contains the following appendix sections: Radiative heat transfer properties for black liquor combustion -- Facilities and techniques and Spectral absorbance and emittance data; and Radiate heat transfer determination of the optical constants of ash samples from kraft recovery boilers -- Calculation procedure; Computation program; Density determination; Particle diameter determination; Optical constant data; and Uncertainty analysis.

  6. Deep Borehole Disposal of Spent Fuel. (Conference) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Deep Borehole Disposal of Spent Fuel. Citation Details In-Document Search Title: Deep Borehole Disposal of Spent Fuel. Abstract not provided. Authors: Brady, Patrick V. Publication...

  7. Activities Related to Storage of Spent Nuclear Fuel | Department...

    Office of Environmental Management (EM)

    Related to Storage of Spent Nuclear Fuel More Documents & Publications Nuclear Regulatory Commission Fifth National Report for the Joint Convention on the Safety of Spent...

  8. National Report Joint Convention on the Safety of Spent Fuel...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management National Report Joint Convention on the Safety of Spent Fuel Management...

  9. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

  10. Current Status of the Spent Nuclear Fuel Management Program in...

    Office of Scientific and Technical Information (OSTI)

    Current Status of the Spent Nuclear Fuel Management Program in the United States. Citation Details In-Document Search Title: Current Status of the Spent Nuclear Fuel Management...

  11. Recovery of manganese oxides from spent alkaline and zinccarbon batteries. An application as catalysts for VOCs elimination

    SciTech Connect (OSTI)

    Gallegos, Mara V.; Falco, Lorena R.; Peluso, Miguel A.; Sambeth, Jorge E.; Thomas, Horacio J.

    2013-06-15

    Highlights: Manganese oxides were synthesized using spent batteries as raw materials. Spent alkaline and zinccarbon size AA batteries were used. A biohydrometallurgical process was employed to bio-lixiviate batteries. Manganese oxides were active in the oxidation of VOCs (ethanol and heptane). - Abstract: Manganese, in the form of oxide, was recovered from spent alkaline and zinccarbon batteries employing a biohydrometallurgy process, using a pilot plant consisting in: an air-lift bioreactor (containing an acid-reducing medium produced by an Acidithiobacillus thiooxidans bacteria immobilized on elemental sulfur); a leaching reactor (were battery powder is mixed with the acid-reducing medium) and a recovery reactor. Two different manganese oxides were recovered from the leachate liquor: one of them by electrolysis (EMO) and the other by a chemical precipitation with KMnO{sub 4} solution (CMO). The non-leached solid residue was also studied (RMO). The solids were compared with a MnO{sub x} synthesized in our laboratory. The characterization by XRD, FTIR and XPS reveal the presence of Mn{sub 2}O{sub 3} in the EMO and the CMO samples, together with some Mn{sup 4+} cations. In the solid not extracted by acidic leaching (RMO) the main phase detected was Mn{sub 3}O{sub 4}. The catalytic performance of the oxides was studied in the complete oxidation of ethanol and heptane. Complete conversion of ethanol occurs at 200 C, while heptane requires more than 400 C. The CMO has the highest oxide selectivity to CO{sub 2}. The results show that manganese oxides obtained using spent alkaline and zinccarbon batteries as raw materials, have an interesting performance as catalysts for elimination of VOCs.

  12. HIGHLY ENERGY EFFICIENT D-GLU (DIRECTED-GREEN LIQ-UOR UTILIZATION) PULPING

    SciTech Connect (OSTI)

    Lucia, Lucian A

    2013-04-19

    Purpose: The purpose of the project was to retrofit the front end (pulp house) of a commercial kraft pulping mill to accommodate a mill green liquor (GL) impregna-tion/soak/exposure and accrue downstream physical and chemical benefits while prin-cipally reducing the energy footprint of the mill. A major player in the mill contrib-uting to excessive energy costs is the lime kiln. The project was intended to offload the energy (oil or natural gas) demands of the kiln by by-passing the causticization/slaking site in the recovery area and directly using green liquor as a pulping medium for wood. Scope: The project was run in two distinct, yet mutually compatible, phases: Phase 1 was the pre-commercial or laboratory phase in which NC State University and the Insti-tute of Paper Science and Technology (at the Georgia Institute of Technology) ran the pulping and associated experiments, while Phase 2 was the mill scale trial. The first tri-al was run at the now defunct Evergreen Pulp Mill in Samoa, CA and lead to a partial retrofit of the mill that was not completed because it went bankrupt and the work was no longer the low-hanging fruit on the tree for the new management. The second trial was run at the MeadWestvaco Pulp Mill in Evedale, TX which for all intents and pur-poses was a success. They were able to fully retrofit the mill, ran the trial, studied the pulp properties, and gave us conclusions.

  13. Spent Nuclear Fuel Project Safety Management Plan

    SciTech Connect (OSTI)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities.

  14. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    SciTech Connect (OSTI)

    Kaplan, P.; Vinoche, R.; Devezeaux, J-G.; Bailly, F.

    2003-02-25

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal.

  15. Pyrochemical processing of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Laidler, J.J.

    1995-02-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

  16. Thermal reclamation of spent blasting abrasive

    SciTech Connect (OSTI)

    Bryan, B.G. ); Thomas, W.; Adema, C. )

    1990-01-01

    Abrasive blasting media is used to remove anticorrosive and antifoulant coatings from the hulls and tanks of US Navy ships. The total production of paint-contaminated spent abrasives from the eight US. Navy shipyards ranges from 75,000 to 100,000 tons per year. Most of this spent abrasive is disposed in landfills. Organic paint binders and heavy metals are present in the spent abrasives in concentrations sufficient to classify them as hazardous wastes in some states. In an effort to avoid the rising costs an long-term environmental liability associated with landfilling this waste, the US Navy has investigated various methods of reclaiming spent abrasives for reuse in hull- and tank-blasting operations. This paper discusses the results of a research and development project conducted under the Navy's Hazardous Waste Minimization Program to test a fluidized-bed sloped-grid (FBSG) reclaimer to determine if it could be used to recycle spent abrasive. Thirty tons of abrasive were processed and a product meeting military specifications for new abrasives was reclaimed. Blasting performance was also comparable to new abrasives. 3 refs., 1 fig., 2 tabs.

  17. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect (OSTI)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  18. Spent Fuel Background Report Volume I

    SciTech Connect (OSTI)

    Abbott, D.

    1994-03-01

    This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic repository and the decision to phase out reprocessing of production fuels are extending the need for interim storage. The report describes the basic storage conditions and the general SNF inventory at individual DOE facilities.

  19. Apparatus for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  20. Method for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  1. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (OSTI)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore » of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  2. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (OSTI)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore »of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  3. Numerical Estimation of the Spent Fuel Ratio

    SciTech Connect (OSTI)

    Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason; Margraf, J.; Dunn, T. A.

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  4. Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (OSTI)

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  5. EM Completes Project to Maintain Water Quality of Spent Nuclear...

    Energy Savers [EERE]

    Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site EM Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site ...

  6. EM Safely and Efficiently Manages Spent Nuclear Fuel | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    (MVDS) is a dry storage system used at Idaho Engineering Laboratory and at Fort St. Vrain to store spent nuclear fuel. Each storage location holds one spent nuclear container. ...

  7. Report on interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  8. January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources |

    Energy Savers [EERE]

    Department of Energy January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - January 28, 2016 - Borehole Disposal of Spent Radioactive Sources (Dr. Matt Kozak, INTERA). Webinar Recording PDF icon Agenda & Webinar Instructions - January 28, 2016 - P&RA CoP Webinar PDF icon Borehole Disposal of Spent Sources

  9. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp

  10. Naval Spent Fuel Rail Shipment Accident Exercise Objectives | Department of

    Office of Environmental Management (EM)

    Energy Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives PDF icon Naval Spent Fuel Rail Shipment Accident Exercise Objectives More Documents & Publications TEC Meeting Summaries - April 2005 Presentations TEC Meeting Summaries - January - February 2007 Presentations NTSF 2014 Meeting Agenda

  11. Impact analysis of spent fuel jacket assemblies

    SciTech Connect (OSTI)

    Aramayo, G.A.

    1994-06-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered.

  12. Spent Sealed Sources Management in Switzerland - 12011

    SciTech Connect (OSTI)

    Beer, H.F.

    2012-07-01

    Information is provided about the international recommendations for the safe management of disused and spent sealed radioactive sources wherein the return to the supplier or manufacturer is encouraged for large radioactive sources. The legal situation in Switzerland is described mentioning the demand of minimization of radioactive waste as well as the situation with respect to the interim storage facility at the Paul Scherrer Institute (PSI). Based on this information and on the market situation with a shortage of some medical radionuclides the management of spent sealed sources is provided. The sources are sorted according to their activity in relation to the nuclide-specific A2-value and either recycled as in the case of high active sources or conditioned as in the case for sources with lower activity. The results are presented as comparison between recycled and conditioned activity for three selected nuclides, i.e. Cs-137, Co-60 and Am-241. (author)

  13. Spent fuel container alignment device and method

    DOE Patents [OSTI]

    Jones, Stewart D.; Chapek, George V.

    1996-01-01

    An alignment device is used with a spent fuel shipping container including a plurality of fuel pockets for spent fuel arranged in an annular array and having a rotatable cover including an access opening therein. The alignment device includes a lightweight plate which is installed over the access opening of the cover. A laser device is mounted on the plate so as to emit a laser beam through a laser admittance window in the cover into the container in the direction of a pre-established target associated with a particular fuel pocket. An indexing arrangement on the container provides an indication of the angular position of the rotatable cover when the laser beam produced by the laser is brought into alignment with the target of the associated fuel pocket.

  14. EIS-0015: U.S. Spent Fuel Policy

    Broader source: Energy.gov [DOE]

    Subsumed DOE/EIS-0040 and DOE/EIS-0041. The Savannah River Laboratory prepared this EIS to analyze the impacts of implementing or not implementing the policy for interim storage of spent power reactor fuel. This Final EIS is a compilation of three Draft EISs and one Supplemental Draft EIS: DOE/EIS-0015-D, Storage of U.S. Spent Power Reactor Fuel; DOE/EIS-0015-DS, Storage of U.S. Spent Power Reactor Fuel - Supplement; DOE/EIS-0040-D, Storage of Foreign Spent Power Reactor Fuel; and DOE/EIS-0041-D, Charge for Spent Fuel Storage.

  15. Spent fuel pool analysis using TRACE code

    SciTech Connect (OSTI)

    Sanchez-Saez, F.; Carlos, S.; Villanueva, J. F.; Martorell, S.

    2012-07-01

    The storage requirements of Spent Fuel Pools have been analyzed with the purpose to increase their rack capacities. In the past, the thermal limits have been mainly evaluated with conservative codes developed for this purpose, although some works can be found in which a best estimate code is used. The use of best estimate codes is interesting as they provide more realistic calculations and they have the capability of analyzing a wide range of transients that could affect the Spent Fuel Pool. Two of the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays, TRACE code is being developed to make use of the more favorable characteristics of RELAP-5 and TRAC codes. Among the components coded in TRACE that can be used to construct the model, it is interesting to use the VESSEL component, which has the capacity of reproducing three dimensional phenomena. In this work, a thermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE code is developed. Such model has been used to perform a licensing calculation and the results obtained have been compared with experimental measurements made at the pool, showing a good agreement between the calculations predicted by TRACE and the experimental data. (authors)

  16. Black liquor combustion validated recovery boiler modeling: Final year report. Volume 2 (Appendices I, section 5 and II, section 1)

    SciTech Connect (OSTI)

    Grace, T.M.; Frederick, W.J.; Salcudean, M.; Wessel, R.A.

    1998-08-01

    This project was initiated in October 1990, with the objective of developing and validating a new computer model of a recovery boiler furnace using a computational fluid dynamics (CFD) code specifically tailored to the requirements for solving recovery boiler flows, and using improved submodels for black liquor combustion based on continued laboratory fundamental studies. The key tasks to be accomplished were as follows: (1) Complete the development of enhanced furnace models that have the capability to accurately predict carryover, emissions behavior, dust concentrations, gas temperatures, and wall heat fluxes. (2) Validate the enhanced furnace models, so that users can have confidence in the predicted results. (3) Obtain fundamental information on aerosol formation, deposition, and hardening so as to develop the knowledge base needed to relate furnace model outputs to plugging and fouling in the convective sections of the boiler. (4) Facilitate the transfer of codes, black liquid submodels, and fundamental knowledge to the US kraft pulp industry. Volume 2 contains the last section of Appendix I, Radiative heat transfer in kraft recovery boilers, and the first section of Appendix II, The effect of temperature and residence time on the distribution of carbon, sulfur, and nitrogen between gaseous and condensed phase products from low temperature pyrolysis of kraft black liquor.

  17. Pyrochemical Treatment of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

    2005-10-01

    Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energys Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

  18. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting the DOE waste canisters and their contents from damage/degradation by the external environment. The disposal containers also interface with the SNF by limiting access of moderator and oxidizing agents to the waste. The disposal containers interface with the Ex-Container System's emplacement drift disposal container supports. The disposal containers interface with the Canister Transfer System, Waste Emplacement System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and remediation of the disposal container.

  19. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  20. Spent fuel management fee methodology and computer code user's manual.

    SciTech Connect (OSTI)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively.

  1. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Lessons Learned Report, NNSA, Feb 2010 | Department of Energy Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 PDF icon 000466 Idaho Spent Fuel Facility (ISFF) Project Appropriate Acquisition Strategy Lessons Learned Report Feb 2011.pdf More Documents & Publications Integrated Biorefinery Research Facility (IBRF

  2. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing

    Energy Savers [EERE]

    U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany | Department of Energy 77: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany SUMMARY This EA will evaluate the potential environmental impacts of a DOE proposal to accept spent nuclear fuel from the

  3. EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department

    Office of Environmental Management (EM)

    of Energy 79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina SUMMARY The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. PUBLIC COMMENT

  4. Pyroprocess for processing spent nuclear fuel

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

    This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

  5. Systems impacts of spent fuel disassembly alternatives

    SciTech Connect (OSTI)

    Not Available

    1984-07-01

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables.

  6. Evaluation of Options for Permanent Geologic Disposal of Spent...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level ...

  7. Management of Spent Dessicant from Vapour Recovery Dryers

    Office of Environmental Management (EM)

    *Significant literature exists on the performance characteristics and on tritium "memory" effects *Focus of request is more on the disposal of spent molecular sieves Purpose...

  8. Deep Borehole Disposal of Spent Fuel. Brady, Patrick V. Abstract...

    Office of Scientific and Technical Information (OSTI)

    Spent Fuel. Brady, Patrick V. Abstract not provided. Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States) USDOE National Nuclear Security Administration (NNSA)...

  9. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  10. US Spent (Used) Fuel Status, Management and Likely Directions- 12522

    SciTech Connect (OSTI)

    Jardine, Leslie J.

    2012-07-01

    As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)

  11. Huizenga leads safety of spent fuel management, radioactive waste...

    National Nuclear Security Administration (NNSA)

    Huizenga leads safety of spent fuel management, radioactive waste management meeting in Vienna | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People...

  12. Production of ammonium sulfate fertilizer from FGD waste liquors. Quarterly technical report, October 1, 1994--December 31, 1994

    SciTech Connect (OSTI)

    Randolph, A.D.; Mukhopadhyay, S.; Unrau, E.

    1994-12-31

    During this quarterly period, an experimental investigation was performed to study the precipitation kinetics and hydrolysis characteristics of calcium imido disulfonate crystals (CaADS). The CaADS crystals were precipitated by a metathetical reaction of lime, supplied by Dravo Lime Co., with flue gas desulfurization (FGD) scrubber waste liquor. Before approaching for the continuous Double Draw-Off (DDO) crystallization studies, the influence of a Dravo lime slurry on the precipitation characteristics of N-S compounds will be established. A series of N-S compound batch crystallization studies were completed in a wide range of pH (7.0--9.0), and the influence of pH on the amount of lime required, as well as the amount of precipitate obtained, was investigated. Although the amount of precipitate increased with increase in solution pH, the safe or optimum pH for the precipitation of CaADS lies in the vicinity of 8.2 to 8.3. For studying the crystallization characteristics of CaADS crystals, a bench scale 7.0 liter DDO crystallizer was built. DDO crystallizer is found to be superior compared to Mixed Suspension Mixed Product Removal (MSMPR) crystallizer. The precipitated crystals were analyzed for elemental composition by chemical analysis. The crystals were also examined under optical microscope for their morphological features. The present studies confirmed our prediction that N-S compounds in the waste liquor can be precipitated by a reaction with lime slurry. The precipitated crystals were mostly calcium imido disulfonate.

  13. Regeneration of FGD waste liquors: Production of ammonium and potassium sulfate mixed fertilizer. Quarterly technical report, April 1993--June 1993

    SciTech Connect (OSTI)

    Randolph, A.D.; Kwon, T.M.

    1993-12-01

    Precipitation and hydrolysis of the N-S compounds in the waste scrubbing liquor provided by Dravo Lime Co. was investigated. Precipitation of N-S compounds by a metathetical reaction with potassium sulfate was performed in continuous crystallizers. A preliminary operation showed that compared to a typical Mixed-Suspension-Mixed-Product-Removal (MSMPR) crystallizer, the Double-Draw-Off (DDO) crystallizer was superior by increasing the average size of the precipitated crystals of N-S compounds from 173 {mu}m to 218 {mu}m. However, the hydrolysis characteristics of the precipitated crystals were not dependent upon crystallizer type. A brief description of a new process which uses lime/limestone for precipitation of N-S compounds in the scrubbing liquor is presented. Preliminary investigations showed the lime/limestone process is efficient in precipitating N-S compounds and the precipitated crystals were shown to be more easily hydrolyzed than potassium salts of N-S compounds. This lime/limestone process is a novel process which seems better than the K{sub 2}SO{sub 4} process because one does not need to purchase/introduce a new chemical additive to precipitate in the lime/limestone Fe-EDTA wet scrubbing processes. Up to the present, the authors focused on developing the K{sub 2}SO{sub 4} process following their original proposal. However, the new lime/limestone process seems more advantageous in terms of economy and environmental safety. Therefore, it seems desirable changing research phase and putting an emphasis on the development of the lime/limestone process. Future study will include investigation of the DDO crystallizer operation to increase the size of precipitated crystals and thus to enhance their processibility. This study seems to be essential to the new lime/limestone process since the precipitated crystals are relatively small in size and thus poor in filterability.

  14. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect (OSTI)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  15. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering

    Office of Environmental Management (EM)

    Laboratory Environmental Restoration and Waste Management Programs | Department of Energy 3: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and

  16. Arrival condition of spent fuel after storage, handling, and transportation

    SciTech Connect (OSTI)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

  17. Technical bases for interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.

    1981-06-01

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be.

  18. HTGR Spent Fuel Treatment Program. HTGR Spent Fuel Treatment Development Program Plan

    SciTech Connect (OSTI)

    Not Available

    1984-12-01

    The spent fuel treatment (SFT) program plan addresses spent fuel volume reduction, packaging, storage, transportation, fuel recovery, and disposal to meet the needs of the HTGR Lead Plant and follow-on plants. In the near term, fuel refabrication will be addressed by following developments in fresh fuel fabrication and will be developed in the long term as decisions on the alternatives dictate. The formulation of this revised program plan considered the implications of the Nuclear Waste Policy Act of 1982 (NWPA) which, for the first time, established a definitive national policy for management and disposal of nuclear wastes. Although the primary intent of the program is to address technical issues, the divergence between commercial and government interests, which arises as a result of certain provisions of the NWPA, must be addressed in the economic assessment of technically feasible alternative paths in the management of spent HTGR fuel and waste. This new SFT program plan also incorporates a significant cooperative research and development program between the United States and the Federal Republic of Germany. The major objective of this international program is to reduce costs by avoiding duplicate efforts.

  19. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    SciTech Connect (OSTI)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

  20. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect (OSTI)

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  1. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect (OSTI)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of contamination. The thermal decomposition of this material is also undesirable if the cladding hulls are melted for volume reduction or to produce waste forms. Handling and disposal of the corrosive off-gas stream and ZrO{sub 2}-containing dross must be addressed. The stability of Zr{sup 4+} in the NHF{sub 4}/NH{sub 4}NO{sub 3} solution is also a concern. Precipitation of ammonium zirconium fluorides upon cooling of the dissolving solution was observed in the feasibility experiments. Precipitation of the solids was attributed to the high fluoride to Zr ratios used in the experiments. The solubility of Zr{sup 4+} in NH{sub 4}F solutions decreases as the free fluoride concentration increases. The removal of the ZrO{sub 2} layer from Zircaloy-4 coupons with HF showed a strong dependence on both the concentration and temperature. Very rapid dissolution of the oxide layer and significant amounts of metal was observed in experiments using HF concentrations {ge} 2.5 M. Treatment of the coupons using HF concentrations {le} 1.0 M was very effective in removing the oxide layer. The most effective conditions resulted in dissolution rates which were less than approximately 2 mg/cm{sup 2}-min. With dissolution rates in this range, uniform removal of the oxide layer was obtained and a minimal amount of Zircaloy metal was dissolved. Future HF dissolution studies should focus on the decontamination of actual spent fuel cladding hulls to determine if the treated hulls meet criteria for disposal as a LLW.

  2. U.S. Spent Nuclear Fuel Data as of December 31, 2002

    U.S. Energy Information Administration (EIA) Indexed Site

    Latest spent nuclear data Release Date: October 1, 2004 Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive...

  3. Loss of spent fuel pool cooling PRA: Model and results

    SciTech Connect (OSTI)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  4. Method of cleaning a spent fuel assembly

    SciTech Connect (OSTI)

    Chung, D.K.; Jones, C.E. Jr.

    1989-05-09

    A method is described of cleaning a fuel assembly including surfaces thereof prior to decladding, each assembly surface contaminated with a radioactive alkali metal and comprising a plurality of pressurized metallic fuel pins containing a spent fissible material, the method comprising the sequential steps of: (a) placing the fuel assembly in a sealed chamber; (b) passing a heated, inert gas through the chamber to heat the fuel assembly to a temperature sufficient to cause volatilization of the alkali metal but insufficient to rupture the pressurized metal pins; (c) evacuating the chamber to a pressure of less than 0.5 mm of Hg to further enhance volatilization and removal of the alkali metal and maintaining the chamber at that pressure until the decay heat of the fissile materials causes the temperature of the fuel assembly to increase to a level which would be detrimental to the integrity of the metal pins; (d) cooling the fuel assembly by passing a cool, inert gas through the chamber to reduce the temperature of the fuel assembly to a desired level; (e) repeating the evacuation and cooling steps as required to insure removal of substantially all of the radioactive alkali metal from the assembly surface; and (f) recovering the cleaned fuel assembly from the chamber.

  5. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

  6. Spent nuclear fuel project path forward preliminary safety evaluation

    SciTech Connect (OSTI)

    Brehm, J.R.; Crowe, R.D.; Siemer, J.M.; Wojdac, L.F.; Hosler, A.G.

    1995-03-01

    This preliminary safety evaluation (PSE) provides validation of the initial project design criteria for the Spent Nuclear Fuel Project (SNFP) Path Forward for removal of fuel from K Basins.

  7. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched ...

  8. Black liquor combustion validated recovery boiler modeling: Final year report. Volume 1 (Main text and Appendix I, sections 1--4)

    SciTech Connect (OSTI)

    Grace, T.M.; Frederick, W.J.; Salcudean, M.; Wessel, R.A.

    1998-08-01

    This project was initiated in October 1990, with the objective of developing and validating a new computer model of a recovery boiler furnace using a computational fluid dynamics (CFD) code specifically tailored to the requirements for solving recovery boiler flows, and using improved submodels for black liquor combustion based on continued laboratory fundamental studies. The key tasks to be accomplished were as follows: (1) Complete the development of enhanced furnace models that have the capability to accurately predict carryover, emissions behavior, dust concentrations, gas temperatures, and wall heat fluxes. (2) Validate the enhanced furnace models, so that users can have confidence in the predicted results. (3) Obtain fundamental information on aerosol formation, deposition, and hardening so as to develop the knowledge base needed to relate furnace model outputs to plugging and fouling in the convective sections of the boiler. (4) Facilitate the transfer of codes, black liquid submodels, and fundamental knowledge to the US kraft pulp industry. Volume 1 contains the main body of the report and the first 4 sections of Appendix 1: Modeling of black liquor recovery boilers -- summary report; Flow and heat transfer modeling in the upper furnace of a kraft recovery boiler; Numerical simulation of black liquor combustion; and Investigation of turbulence models and prediction of swirling flows for kraft recovery furnaces.

  9. Extraction of uranium from spent fuels using liquefied gases

    SciTech Connect (OSTI)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2007-07-01

    For reprocessing of spent nuclear fuels, a novel method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. As a fundamental study, the nitrate conversion with liquefied nitrogen dioxide and the nitrate extraction with supercritical carbon dioxide were demonstrated by using uranium dioxide powder, uranyl nitrate and tri-n-butylphosphate complex in the present study. (authors)

  10. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    SciTech Connect (OSTI)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post loss of water inventory hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  11. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  12. Safeguards Guidance for Independent Spent Fuel Storage Installations (ISFSI)

    National Nuclear Security Administration (NNSA)

    1) May 2012 Guidance for Independent Spent Fuel Dry Storage Installations Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI) Philip Casey Durst (INL Consultant) DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy,

  13. The united kingdom's changing requirements for spent fuel storage

    SciTech Connect (OSTI)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N.

    2013-07-01

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK.

  14. EDI as a Treatment Module in Recycling Spent Rinse Waters

    SciTech Connect (OSTI)

    Donovan, Robert P.; Morrison, Dennis J.

    1999-08-11

    Recycling of the spent rinse water discharged from the wet benches commonly used in semiconductor processing is one tactic for responding to the targets for water usage published in the 1997 National Technology Roadmap for Semiconductors (NTRS). Not only does the NTRS list a target that dramatically reduces total water usage/unit area of silicon manufactured by the industry in the future but for the years 2003 and beyond, the NTRS actually touts goals which would have semiconductor manufacturers drawing less water from a regional water supply per unit area of silicon manufactured than the quantity of ultrapure water (UPW) used in the production of that same silicon. Achieving this latter NTRS target strongly implies more widespread recycling of spent rinse waters at semiconductor manufacturing sites. In spite of the fact that, by most metrics, spent rinse waters are of much higher purity than incoming municipal waters, recycling of these spent rinse waters back into the UPW production plant is not a simple, straightforward task. The rub is that certain of the chemicals used in semiconductor manufacturing, and thus potentially present in trace concentrations (or more) in spent rinse waters, are not found in municipal water supplies and are not necessarily removed by the conventional UPW production sequence used by semiconductor manufacturers. Some of these contaminants, unique to spent rinse waters, may actually foul the resins and membranes of the UPW system, posing a threat to UPW production and potentially even causing a shutdown.

  15. Black liquor combustion validated recovery boiler modeling: Final year report. Volume 3 (Appendices II, sections 2--3 and III)

    SciTech Connect (OSTI)

    Grace, T.M.; Frederick, W.J.; Salcudean, M.; Wessel, R.A.

    1998-08-01

    This project was initiated in October 1990, with the objective of developing and validating a new computer model of a recovery boiler furnace using a computational fluid dynamics (CFD) code specifically tailored to the requirements for solving recovery boiler flows, and using improved submodels for black liquor combustion based on continued laboratory fundamental studies. The key tasks to be accomplished were as follows: (1) Complete the development of enhanced furnace models that have the capability to accurately predict carryover, emissions behavior, dust concentrations, gas temperatures, and wall heat fluxes. (2) Validate the enhanced furnace models, so that users can have confidence in the predicted results. (3) Obtain fundamental information on aerosol formation, deposition, and hardening so as to develop the knowledge base needed to relate furnace model outputs to plugging and fouling in the convective sections of the boiler. (4) Facilitate the transfer of codes, black liquid submodels, and fundamental knowledge to the US kraft pulp industry. Volume 3 contains the following appendix sections: Formation and destruction of nitrogen oxides in recovery boilers; Sintering and densification of recovery boiler deposits laboratory data and a rate model; and Experimental data on rates of particulate formation during char bed burning.

  16. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    SciTech Connect (OSTI)

    Botsch, W.; Smalian, S.; Hinterding, P.

    2013-07-01

    Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  17. Historical overview of domestic spent fuel shipments: Update

    SciTech Connect (OSTI)

    Not Available

    1991-07-01

    This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate that the greatest number of research reactor fuel shipments occurred in 1986. The largest campaigns in 1986 were from the Brookhaven National Laboratory, Brooklyn, New York, to the Idaho Chemical Processing Plant (ICPP) and from the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in Tennessee and the Rockwell International Reactor in California to the Savannah River Plant near Aiken, South Carolina. For all years addressed in this report, DOE facilities in Idaho Falls and Savannah River were the major recipients of research reactor spent fuel. In 1989, 10 shipments were received at the Idaho facilities. These originated from universities in California, Michigan, and Missouri. 9 refs., 12 figs., 7 tabs.

  18. Mission Need Statement: Idaho Spent Fuel Facility Project

    SciTech Connect (OSTI)

    Barbara Beller

    2007-09-01

    Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

  19. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  20. Information handbook on independent spent fuel storage installations

    SciTech Connect (OSTI)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  1. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.

  2. NUHOMS modular spent-fuel storage system: Performance testing

    SciTech Connect (OSTI)

    Strope, L.A.; McKinnon, M.A. ); Dyksterhouse, D.J.; McLean, J.C. )

    1990-09-01

    This report documents the results of a heat transfer and shielding performance evaluation of the NUTECH HOrizontal MOdular Storage (NUHOMS{reg sign}) System utilized by the Carolina Power and Light Co. (CP L) in an Independent Spent Fuel Storage Installation (ISFSI) licensed by the US Nuclear Regulatory Commission (NRC). The ISFSI is located at CP L's H. B. Robinson Nuclear Plant (HBR) near Hartsville, South Carolina. The demonstration included testing of three modules, first with electric heaters and then with spent fuel. The results indicated that the system was conservatively designed, with all heat transfer and shielding design criteria easily met. 5 refs., 45 figs., 9 tabs.

  3. Method for the regeneration of spent molten zinc chloride

    DOE Patents [OSTI]

    Zielke, Clyde W.; Rosenhoover, William A.

    1981-01-01

    In a process for regenerating spent molten zinc chloride which has been used in the hydrocracking of coal or ash-containing polynuclear aromatic hydrocarbonaceous materials derived therefrom and which contains zinc chloride, zinc oxide, zinc oxide complexes and ash-containing carbonaceous residue, by incinerating the spent molten zinc chloride to vaporize the zinc chloride for subsequent condensation to produce a purified molten zinc chloride: an improvement comprising the use of clay in the incineration zone to suppress the vaporization of metals other than zinc. Optionally water is used in conjunction with the clay to further suppress the vaporization of metals other than zinc.

  4. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect (OSTI)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

  5. Spent fuel dissolution studies FY 1991 to 1994

    SciTech Connect (OSTI)

    Gray, W.J.; Wilson, C.N.

    1995-12-01

    Dissolution and transport as a result of groundwater flow are generally accepted as the primary mechanisms by which radionuclides from spent fuel placed in a geologic repository could be released to the biosphere. To help provide a source term for performance assessment calculations, dissolution studies on spent fuel and unirradiated uranium oxides have been conducted over the past few years at Pacific Northwest National Laboratory (PNNL) in support of the Yucca Mountain Site Characterization Project. This report describes work for fiscal years 1991 through 1994. The objectives of these studies and the associated conclusions, which were based on the limited number of tests conducted so far, are described in the following subsections.

  6. Role of vanadium(V) in the aging of the organic phase in the extraction of uranium(VI) by Alamine 336 from acidic sulfate leach liquors

    SciTech Connect (OSTI)

    Chagnes, A.; Cote, G.; Courtaud, B.; Thiry, J.

    2008-07-01

    The present work is focussed on the chemical degradation of Alamine 336-tridecanol-n-dodecane solvent which used in the recovery of uranium by solvent extraction. Degradation occurs due to the presence of vanadium(V), an oxidant, in the feed solution. After a brief overview of the chemistry of vanadium, the kinetics of degradation of the solvent when contacted with acidic sulfate leach liquor was investigated and interpreted by the Michelis-Menten mechanism. GCMS analyses evidenced the presence of tridecanoic acid and dioctylamine as degradation products. A mechanism of degradation is discussed. (authors)

  7. Separator assembly for use in spent nuclear fuel shipping cask

    DOE Patents [OSTI]

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  8. Pinhole Breaches in Spent Fuel Containers: Some Modeling Considerations

    SciTech Connect (OSTI)

    Casella, Andrew M.; Loyalka, Sudarsham K.; Hanson, Brady D.

    2006-06-04

    This paper replaces PNNL-SA-48024 and incorporates the ANS reviewer's comments, including the change in the title. Numerical methods to solve the equations for gas diffusion through very small breaches in spent fuel containers are presented and compared with previous literature results.

  9. Huizenga leads safety of spent fuel management, radioactive waste

    National Nuclear Security Administration (NNSA)

    management meeting in Vienna | National Nuclear Security Administration Huizenga leads safety of spent fuel management, radioactive waste management meeting in Vienna | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget

  10. President Reagan Calls for a National Spent Fuel Storage Facility |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration Reagan Calls for a National Spent Fuel Storage Facility | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets

  11. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  12. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect (OSTI)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the worlds first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  13. What to Expect When Readying to Move Spent Nuclear Fuel from...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power ...

  14. EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries.  The spent fuel would be shipped across...

  15. Influence of corn steep liquor and glucose on colonization of control and CCB (Cu/Cr/B)-treated wood by brown rot fungi

    SciTech Connect (OSTI)

    Humar, Miha; Pohleven, Franc

    2006-07-01

    There are increasing problems with regard to the disposal of treated wood waste. Due to heavy metals or arsenic in impregnated wood waste, burning and landfill disposal options are not considered to be environmentally friendly solutions for dealing with this problem. Extraction of the heavy metals and recycling of the preservatives from the wood waste is a much more promising and environmentally friendly solution. In order to study the scale up of this process, copper/chromium/boron-treated wood specimens were exposed to copper tolerant (Antrodia vaillantii and Leucogyrophana pinastri) and copper sensitive wood decay fungi (Gloeophyllum trabeum and Poria monticola). Afterwards, the ability of fungal hyphae to penetrate and overgrow the wood specimens was investigated. The fungal growths were stimulated by immersing the specimens into aqueous solution of glucose or corn steep liquor prior to exposure to the fungi. The fastest colonization of the impregnated wood was by the copper tolerant A. vaillantii. Addition of glucose onto the surface of the wood specimens increased the fungi colonization of the specimens; however, immersion of the specimens into the solution of corn steep liquor did not have the same positive influence. These results are important in elucidating copper toxicity in wood decay fungi and for using these fungi for bioremediation of treated wood wastes.

  16. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  17. Removal of arsenic compounds from spent catecholated polymer

    DOE Patents [OSTI]

    Fish, Richard H.

    1985-01-01

    Described is a process for removing arsenic from petroliferous derived liquids by contacting said liquid at an elevated temperature with a divinylbenzene-crosslinked polystyrene having catechol ligands anchored thereon. Also, described is a process for regenerating spent catecholated polystyrene by removal of the arsenic bound to it from contacting petroliferous liquid as described above and involves: a. treating said spent catecholated polystyrene, at a temperature in the range of about 20.degree. to 100.degree. C. with an aqueous solution of at least one carbonate and/or bicarbonate of ammonium, alkali and alkaline earth metals, said solution having a pH between about 8 and 10 and, b. separating the solids and liquids from each other. Preferably the regeneration treatment is in two steps wherein step (a) is carried out with an aqueous alcoholic carbonate solution containing lower alkyl alcohol, and, steps (a) and (b) are repeated using a bicarbonate.

  18. Molten tin reprocessing of spent nuclear fuel elements

    DOE Patents [OSTI]

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  19. Hot startup experience with electrometallurgical treatment of spent nuclear fuel

    SciTech Connect (OSTI)

    Benedict, R.W.; Lineberry, M.J.; McFarlane, H.F.; Rigg, R.H.

    1997-10-01

    The treatment of spent metal fuel from the EBR-II fast reactor commenced in June of 1996 at the Fuel Conditioning Facility on the Argonne-West site in Idaho, USA. During the first year of hot operations, 20 fuel assemblies entered processing and 6 low enrichment uranium product ingots were produced. Results are presented for the various process steps with decontamination factors achieved and equipment operational history reported.

  20. Recycling of nuclear spent fuel with AIROX processing

    SciTech Connect (OSTI)

    Majumdar, D.; Jahshan, S.N.; Allison, C.M.; Kuan, P.; Thomas, T.R.

    1992-12-01

    This report examines the concept of recycling light water reactor (LWR) fuel through use of a dry-processing technique known as the AIROX (Atomics International Reduction Oxidation) process. In this concept, the volatiles and the cladding from spent LWR fuel are separated from the fuel by the AIROX process. The fuel is then reenriched and made into new fuel pins with new cladding. The feasibility of the concept is studied from a technical and high level waste minimization perspective.

  1. Handling encapsulated spent fuel in a geologic repository environment

    SciTech Connect (OSTI)

    Ballou, L.B.

    1983-02-01

    In support of the Spent Fuel Test-Climate at the U.S. Department of Energy`s Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site ({similar_to}100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground.

  2. Adsorption of phenol from aqueous systems onto spent oil shale

    SciTech Connect (OSTI)

    Darwish, N.A.; Halhouli, K.A.; Al-Dhoon, N.M. [Jordan Univ. of Science and Technology, Irbid (Jordan)

    1996-03-01

    To evaluate its ability to remove phenol from aqueous solution, Jordanian {open_quotes}spent{close_quotes} oil shale, an abundant natural resource, has been used in an experimental adsorption study. Equilibrium of the system has been determined at three temperatures: 30, 40, and 55{degrees}C. The resulting experimental equilibrium isotherms are well represented by Frendlich, Langmuir, and Redlich-Peterson isotherms. The relevant parameters for these isotherms, as regressed from the experimental equilibrium data, are presented. Effects of solution pH (in the range of 3-11), in addition to effects of three inorganic salts (Kl, KCl, and NaCl), on the equilibrium isotherms were also investigated. The effects of pH in the presence of KI and NaCl were also investigated for a possible interaction between salts and solution pH. The initial concentration of phenol in the aqueous system studied ranges from 10 to 200 ppm. Experimental results show that while an acidic solution has no effect on the adsorption capacity of spent oil shale to phenol, a highly basic solution reduces its adsorbability. No sound effect was observed for the inorganic salts studied on the adsorption of phenol on spent oil shale. The experimental results show that there is no interaction between the pH of solution and the presence of salts. In spite of its ability to remove phenol, spent oil shale showed a very low equilibrium capacity (of an order of magnitude of 1 mg/g). Should the adsorption capacity of the shale be improved (by different treatment processes, such as grafting, surface conditioning), results of this study will find a direct practical implication in serving as {open_quotes}raw{close_quotes} reference data for comparison purposes.

  3. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management

    Office of Environmental Management (EM)

    NRC's Integrated Strategy for NRC s Integrated Strategy for Spent Fuel Management Earl Easton 1 U.S. Nuclear Regulatory Commission May 25, 2010 Road to Yucca Mountain * 20+ years of preparation for the licensing i review * DOE application received in June 2008 and accepted for review in September 2008 * President Obama pursues alternatives to Yucca Mountain * DOE motion to withdraw in March 2010 2 * DOE motion to withdraw in March 2010 * Blue Ribbon Commission on America's Nuclear Future 2

  4. Method For Processing Spent (Trn,Zr)N Fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Richmann, Michael K. (Woodridge, IL)

    2004-07-27

    A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

  5. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  6. Management of Spent Dessicant from Vapour Recovery Dryers

    Office of Environmental Management (EM)

    Management of Spent Desiccant from Vapour Recovery Dryers Armando Antoniazzi Tritium Focus Group Meeting April 2014 www.kinectrics.com http://blog.kinectrics.com Life Cycle Management Solutions for the Electricity Industry *Request information from members of the tritium handling community *Specifically, practices associated with handling and disposal of tritiated molecular sieves *Use of molecular sieves to trap tritiated water is a familiar technology in the tritium handling community

  7. Thermoelectric powered wireless sensors for spent fuel monitoring

    SciTech Connect (OSTI)

    Carstens, T.; Corradini, M.; Blanchard, J.; Ma, Z.

    2011-07-01

    This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

  8. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    SciTech Connect (OSTI)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  9. Surveillance of LWR spent fuel in wet storage. Final report, October 1984

    SciTech Connect (OSTI)

    Bailey, W.J.; Johnson, A.B. Jr.

    1984-10-01

    Battelle, Pacific Northwest Laboratories established a surveillance program for EPRI that documents the integrity of spent light-water reactor fuel and structural materials (spent fuel storage pool liners, racks, piping, etc.) during wet storage. The program involves providing an update on the overall performance of spent fuel in wet storage, monitoring Licensee Event Reports (LERs) for pertinent significant occurrences, identifying lead spent fuel assemblies that are of particular interest to EPRI, monitoring developments in fuel design and performance and assessing their influence on spent fuel storage characteristics, and identifying specific actions or programs that may be needed to maintain the viability of wet storage of spent fuel for extended periods. Experience to date indicates that wet storage is a well-developed technology with no associated major technological problems. Spent fuel storage pools are operated without substantial risk to the public or the plant personnel. A list of lead spent fuel assemblies is presented. Pertinent occurrences from LERs are listed. Very few fuel assemblies have suffered major mechanical damage as a result of handling operations at spent fuel storage pools. Experience to date with handling operations at spent fuel storage pools indicates that failed fuel rods and inadvertent fracturing of fuel rods can be accommodated. Minor problems have occurred with spent fuel storage pool components such as liners, racks, and piping. Surveillance continues to be needed on: (1) possible effects on handling and storage of spent fuel from extended burnup, hydrogen injection at boiling water reactors, and rod consolidation operations; (2) extended pool exposure of neutron-absorbing materials; (3) cracking of spent fuel storage pool piping at pressurized water reactors; and (4) control of impurities in spent fuel pool waters. 120 references, 13 figures, 10 tables.

  10. An approach to determine a defensible spent fuel ratio.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980's and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000's. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort is urgent in order to secure the only known available SNF samples with a clearly defined path to disposal.

  11. Electrometallurgical treatment of oxide spent fuel - engineering-scale development.

    SciTech Connect (OSTI)

    Karell, E. J.

    1998-04-22

    Argonne National Laboratory (ANL) has developed the electrometallurgical treatment process for conditioning various Department of Energy (DOE) spent fuel types for long-term storage or disposal. This process uses electrorefining to separate the constituents of spent fuel into three product streams: metallic uranium, a metal waste form containing the cladding and noble metal fission products, and a ceramic waste form containing the transuranics, and rare earth, alkali, and alkaline earth fission products. While metallic fuels can be directly introduced into the electrorefiner, the actinide components of oxide fuels must first be reduced to the metallic form. The Chemical Technology Division of AFT has developed a process to reduce the actinide oxides that uses lithium at 650 C in the presence of molten LiCl, yielding the actinide metals and Li{sub 2}O. A significant amount of work has already been accomplished to investigate the basic chemistry of the lithium reduction process and to demonstrate its applicability to the treatment of light-water reactor- (LWR-) type spent fuel. The success of this work has led to conceptual plans to construct a pilot-scale oxide reduction facility at ANL's Idaho site. In support of the design effort, a series of laboratory- and engineering-scale experiments is being conducted using simulated fuel. These experiments have focused on the engineering issues associated with scaling-up the process and proving compatibility between the reduction and electrorefining steps. Specific areas of investigation included reduction reaction kinetics, evaluation of various fuel basket designs, and issues related to electrorefining the reduced product. This paper summarizes the results of these experiments and outlines plans for future work.

  12. Waste management plan for Hanford spent nuclear fuel characterization activities

    SciTech Connect (OSTI)

    Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-17

    A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

  13. APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

    SciTech Connect (OSTI)

    Drayer, R.

    2013-06-09

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each specific need. There are products on the market for smart meters, industrial lighting control and home automation that can be applied to the Back End Fuel Cycle. With a little integration and innovation a cost effective solution is achievable.

  14. Method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  15. Economic Viability of Brewery Spent Grain as a Biofuel

    SciTech Connect (OSTI)

    Morrow, Charles

    2016-01-01

    This report summarizes an investigation into the technical feasibility and economic viability of use grain wastes from the beer brewing process as fuel to generate the heat needed in subsequent brewing process. The study finds that while use of spent grain as a biofuel is technically feasible, the economics are not attractive. Economic viability is limited by the underuse of capital equipment. The investment in heating equipment requires a higher utilization that the client brewer currently anticipates. It may be possible in the future that changing factors may swing the decision to a more positive one.

  16. Method for reprocessing and separating spent nuclear fuels

    DOE Patents [OSTI]

    Krikorian, Oscar H. (Danville, CA); Grens, John Z. (Livermore, CA); Parrish, Sr., William H. (Walnut Creek, CA)

    1983-01-01

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  17. Closure mechanism and method for spent nuclear fuel canisters

    DOE Patents [OSTI]

    Doman, Marvin J. (Monroeville, PA)

    2004-11-23

    A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

  18. SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL

    Energy Savers [EERE]

    SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL TO DOE NATIONAL LABORATORIES FOR RESEARCH AND DEVELOPMENT PURPOSES Office of Nuclear Energy U.S. DEPARTMENT OF ENERGY DECEMBER 2015 DOE/EIS-0203-SA-07 DOE/EIS-0250F-S-1-SA-02 Commercial Fuel Shipment SA DOE/EIS-0203-SA-07 December 2015 CONVERSION FACTORS Metric to English English to Metric Multiply by To get Multiply by To get Area Square kilometers 247.1 Acres Square kilometers 0.3861 Square miles Square meters 10.764 Square

  19. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  20. Technical Development on Burn-up Credit for Spent LWR Fuel

    SciTech Connect (OSTI)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  1. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Decommissioned Reactors | Department of Energy Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors PDF icon Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors More Documents & Publications Information Request, "THE REPORT TO THE PRESIDENT AND THE CONGRESS BY THE SECRETARY OF ENERGY ON THE NEED

  2. National Report Joint Convention on the Safety of Spent Fuel Management and

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    on the Safety of Radioactive Waste Management | Department of Energy National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management This is the first National Report prepared under the terms of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Managementi hereafter

  3. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors

    Energy Savers [EERE]

    6 Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 2008 U.S. Department of Energy Office of Civilian Radioactive Waste Management Washington, D.C. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel The picture on the cover is the Connecticut Yankee Independent Spent Fuel Storage Installation site in Haddam, Connecticut, with 43 dry storage NRC-licensed dual-purpose

  4. Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11

    Energy Savers [EERE]

    Safety Aspects of Wet Storage of Spent Nuclear Fuel OAS-L-13-11 July 2013 Department of Energy Washington, DC 20585 July 10, 2013 MEMORANDUM FOR THE SENIOR ADVISOR FOR ENVIRONMENTAL MANAGEMENT FROM: Daniel M. Weeber Assistant Inspector General for Audits and Administration Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Safety Aspects of Wet Storage of Spent Nuclear Fuel" BACKGROUND The Department of Energy (Department) is responsible for managing and storing spent

  5. Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel

    Office of Environmental Management (EM)

    and High-Level Radioactive Waste | Department of Energy Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring geologic

  6. Spent nuclear fuel recycling with plasma reduction and etching

    DOE Patents [OSTI]

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  7. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect (OSTI)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  8. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect (OSTI)

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  9. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  10. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  11. Spent sealed radium sources conditioning in Latin America

    SciTech Connect (OSTI)

    Mourao, R.P.

    1999-06-01

    The management of spent sealed sources is considered by the International Atomic Energy Agency (IAEA) one of the greatest challenges faced by nuclear authorities today, especially in developing countries. One of the Agency`s initiatives to tackle this problem is the Spent Radium Sources Conditioning Project, a worldwide project relying on the regional co-operation between countries. A team from the Brazilian nuclear research institute Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) was chosen as the expert team to carry out the operations in Latin America; since December 1996 radium sources have been safely conditioned in Uruguay, Nicaragua, Guatemala, Ecuador and Paraguay. A Quality Assurance Program was established, encompassing the qualification of the capsule welding process, written operational procedures referring to all major steps of the operation, calibration of monitors and information retrievability. A 200L carbon steel drum-based packaging concept was used to condition the sources, its cavity being designed to receive the lead shield device containing stainless steel capsules with the radium sources. As a result of these operations, a total amount of 2,897 mg of needles, tubes, medical applicators, standard sources for calibration, lightning rods, secondary wastes and contaminated objects were stored in proper conditions and are now under control of the nuclear authorities of the visited countries.

  12. Development of a thermal reclamation system for spent blasting abrasive

    SciTech Connect (OSTI)

    Bryan, B.B.; Mensinger, M.C.; Rehmat, A.G.

    1991-01-01

    Abrasive blasting is the most economical method for paint removal from large surface areas such as the hulls and tanks of oceangoing vessels. Tens of thousands of tons of spent abrasive are generated annually by blasting operations in private and US Navy shipyards. Some of this material is classified as hazardous waste, and nearly all of it is currently being either stockpiled or disposed in landfills. The rapid decline in available landfill space and corresponding rise in landfill tipping fees pose a severe problem for shipyard operators throughout the US. This paper discusses the results of a research and development program initiated by the Institute of Gas Technology and supported by the US Navy to develop and test a fluidized-bed thermal reclamation system for spent abrasive waste minimization. Bench- and pilot-scale reclaimer tests and reclaimed abrasive performance tests are described along with the current status of a program to build and test a 5-ton/hour prototype reclaimer at a US Navy shipyard.

  13. Report to Congress on Plan for Interim Storage of Spent Nuclear...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors PDF icon Report to Congress on Plan ...

  14. Fact #820: May 5, 2014 Dollars Spent on Imported Petroleum | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energy 0: May 5, 2014 Dollars Spent on Imported Petroleum Fact #820: May 5, 2014 Dollars Spent on Imported Petroleum Over the last three decades, the amount of money the U.S. spent on imported petroleum varied widely. In 1988 and 1998, about $200 million per day was spent on imported petroleum, but in 2008 it was more than $1,300 million per day. In 2009 lower demand and a collapse in oil prices caused the expenditures to fall to about $750 million per day. Since that time the spending has

  15. EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    DOE prepared a EIS that evaluated the potential environmental impacts of treatment and management of DOE-owned sodium bonded spent nuclear fuel.

  16. Oxidative alteration of spent fuel in a silica-rich environment...

    Office of Scientific and Technical Information (OSTI)

    SEMAEM investigation and geochemical modeling Citation Details In-Document Search Title: Oxidative alteration of spent fuel in a silica-rich environment: SEMAEM ...

  17. Locations of Spent Nuclear Fuel and High-Level Radioactive Waste |

    Office of Environmental Management (EM)

    Department of Energy Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Map of the United States of America showing the locations of spent nuclear fuel and high-level radioactive waste. PDF icon Slide 1 More Documents & Publications Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel End of Year 2010 SNF & HLW Inventories Evaluation of Options for Permanent

  18. oint Convention on the Safety of Spent Fuel Management and on...

    National Nuclear Security Administration (NNSA)

    oint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management | National Nuclear Security Administration Facebook Twitter Youtube Flickr...

  19. A Compact Reactor Gate Discharge Monitor for Spent Fuel.

    SciTech Connect (OSTI)

    Franco, J. B.; Menlove, Howard O.; Eccleston, G. W.; Miller, M. C.

    2005-01-01

    This paper presents a new design for a reactor gate discharge monitor that has evolved from the baseline discharge monitors used at the Fugen and Tokai-1 reactors in Japan. The main design innovation is the ability to determine direction-of-motion of spent fuel using a single sensor module, as opposed to the two modules used in both baseline design systems. Use of a single module reduces the final system complexity and weight significantly without compromising functionality. The reactor gate discharge monitor uses standard International Atomic Energy Agency (IAEA) hardware and software components. The requirements to determine direction-of-motion from a single module precipitated several development efforts described herein in both the MiniGRAND data acquisition hardware and in the uninterruptible power supply source.

  20. Training implementation matrix, Spent Nuclear Fuel Project (SNFP)

    SciTech Connect (OSTI)

    EATON, G.L.

    2000-06-08

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

  1. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  2. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (1.4 kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.

    1981-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.4 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a stainless steel canister representative of actual fuel canisters, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel near-surface drywell tests being conducted at E-MAD, the spent fuel deep geologic storage test being conducted in Climax granite on the Nevada Test Site, and for five constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  3. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    SciTech Connect (OSTI)

    Connors, B. J.; Golden, M. P.; Valenti, P. J.; Winkel, J. J.

    1987-03-01

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs.

  4. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  5. The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling,

    Office of Environmental Management (EM)

    Storage, Transportation, and Disposal in the U.S. | Department of Energy The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. PDF icon The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. More Documents & Publications

  6. Comparison of model predictions with measurements using the improved spent fuel attribute tester

    SciTech Connect (OSTI)

    Dupree, S.A.; Laub, T.W.; Arlt, R.

    1994-08-01

    Design improvements for the International Atomic Energy Agency`s Spent Fuel Attribute Tester, recommended on the basis of an optimization study, were incorporated into a new instrument fabricated under the Finnish Support Programme. The new instrument was tested at a spent fuel storage pool on September 8 and 9, 1993. The result of two of the measurements have been compared with calculations. In both cases the calculated and measured pulse height spectra in good agreement and the {sup 137}Cs gamma peak signature from the target spent fuel element is present.

  7. Shipper/receiver difference verification of spent fuel by use of PDET

    SciTech Connect (OSTI)

    Ham, Y. S.; Sitaraman, S.

    2011-07-01

    Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

  8. Long-term management of high-level radioactive waste (HLW) and spent

    Energy Savers [EERE]

    nuclear fuel (SNF) | Department of Energy Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor to generate nuclear energy but that has been removed from the reactor as no

  9. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    SciTech Connect (OSTI)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling).

  10. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S.

    2008-07-01

    This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

  11. A NOVEL APPROACH TO SPENT FUEL POOL DECOMMISSIONING

    SciTech Connect (OSTI)

    R. L. Demmer

    2011-04-01

    The Idaho National Laboratory (INL) has been at the forefront of developing methods to reduce the cost and schedule of deactivating spent fuel pools (SFP). Several pools have been deactivated at the INL using an underwater approach with divers. These projects provided a basis for the INL cooperation with the Dresden Nuclear Power Station Unit 1 SFP (Exelon Generation Company) deactivation. It represents the first time that a commercial nuclear power plant (NPP) SFP was decommissioned using this underwater coating process. This approach has advantages in many aspects, particularly in reducing airborne contamination and allowing safer, more cost effective deactivation. The INL pioneered underwater coating process was used to decommission three SFPs with a total combined pool volume of over 900,000 gallons. INL provided engineering support and shared project plans to successfully initiate the Dresden project. This report outlines the steps taken by INL and Exelon to decommission SFPs using the underwater coating process. The rationale used to select the underwater coating process and the advantages and disadvantages are described. Special circumstances are also discussed, such as the use of a remotely-operated underwater vehicle to visually and radiologically map the pool areas that were not readily accessible. A larger project, the INTEC-603 SFP in-situ (grouting) deactivation, is reviewed. Several specific areas where special equipment was employed are discussed and a Lessons Learned evaluation is included.

  12. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

  13. Accelerator-driven transmutation of spent fuel elements

    DOE Patents [OSTI]

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  14. Table 3. Annual commercial spent fuel discharges and burnup

    Gasoline and Diesel Fuel Update (EIA)

    Next release date: Late 2018 Table 3. Annual commercial spent fuel discharges and burnup, 1968 - June 30, 2013 Number of assemblies a Initial uranium content (MTU) Average burnup (GWDt/MTU) All discharged assemblies Year BWR PWR Total BWR PWR Total BWR PWR 1968 0 0 0 0.0 0.0 0.0 0.0 0.0 1969 94 0 94 9.6 0.0 9.6 16.7 0.0 1970 0 0 0 0.0 0.0 0.0 0.0 0.0 1971 193 25 218 22.5 10.4 32.9 15.9 26.5 1972 256 251 507 42.8 88.0 130.8 12.4 23.3 1973 526 113 639 87.7 48.3 135.9 13.4 21.5 1974 1,166 571 1,737

  15. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    SciTech Connect (OSTI)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  16. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation Gauld, Ian C. Oak Ridge National Lab. (ORNL), Oak Ridge,...

  17. Design criteria for an independent spent fuel storage installation (water pool type)

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    This standard is intended to be used by those involved in the ownership and operation of an Independent Spent Fuel Storage Installation (ISFSI) in specifying the design requirements and by the designer in meeting the minimum design requirements of such installations. This standard continues the set of American National Standards on spent fuel storage design. Similar standards are: Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, N210-1976 (ANS-57.2); Design Objectives for Highly Radioactive Solid Material Handling and Storage Facilities in a Reprocessing Plant, ANSI N305-1975; and Guidelines for Evaluating Site-Related Parameters for an Independent Spent Fuel Storage Installation, ANSI/ANS-2.19-1981.

  18. EM Prepares Report for Convention on Safety of Spent Fuel and...

    Office of Environmental Management (EM)

    The Convention was established in 2001 as the first instrument to directly address issues related to the safety of spent fuel and radioactive waste on a global scale. EM...

  19. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  20. Study of Fukushima Dai-ichi Nuclear Power Station Unit 4 Spent...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Study of Fukushima Dai-ichi Nuclear Power Station Unit 4 Spent Fuel Pool Citation Details In-Document Search Title: Study of Fukushima Dai-ichi Nuclear Power...

  1. EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

  2. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOE Patents [OSTI]

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  3. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    DOE PAGES Search Results Accepted Manuscript: Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation This...

  4. EM Prepares Report for Convention on Safety of Spent Fuel and Radioactive Waste Management

    Broader source: Energy.gov [DOE]

    WASHINGTON, D.C. EM supported DOE in its role as the lead technical agency to produce a report recently for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management.

  5. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    SciTech Connect (OSTI)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  6. Second National Report for the Joint Convention on the Safety of Spent Fuel

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Management and on the Safety of Radioactive Waste Management | Department of Energy Second National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management Second National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management This second National Report updates the first National Report published on May 3, 2003, under the terms of the Joint Convention on the

  7. Oxidative alteration of spent fuel in a silica-rich environment: SEM/AEM

    Office of Scientific and Technical Information (OSTI)

    investigation and geochemical modeling (Conference) | SciTech Connect Oxidative alteration of spent fuel in a silica-rich environment: SEM/AEM investigation and geochemical modeling Citation Details In-Document Search Title: Oxidative alteration of spent fuel in a silica-rich environment: SEM/AEM investigation and geochemical modeling Correctly identifying the possible alteration products and accurately predicting their occurrence in a repository-relevant environment are the key for the

  8. Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY12 Status Report

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY12 Status Report Citation Details In-Document Search Title: Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY12 Status Report Executive Summary The Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign is supporting a multi-institutional collaboration to study the feasibility of using Lead Slowing Down Spectroscopy (LSDS) to conduct direct,

  9. EIS-0453: Recapitalization of Infrastructure Supporting Naval Spent Nuclear Fuel Handling at the Idaho National Laboratory

    Broader source: Energy.gov [DOE]

    The Draft EIS evaluates the potential environmental impacts associated with recapitalizing the infrastructure needed to ensure the long-term capability of the Naval Nuclear Propulsion Program (NNPP) to support naval spent nuclear fuel handling capabilities provided by the Expended Core Facility (ECF). Significant upgrades are necessary to ECF infrastructure and water pools to continue safe and environmentally responsible naval spent nuclear fuel handling until at least 2060.

  10. Third National Report for the Joint Convention on the Safety of Spent Fuel

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Management and on the Safety of Radioactive Waste Management | Department of Energy Third National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management Third National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management This Third United States National Report updates the second National Report published in October 2005, under the terms of the Joint

  11. Spray-dryer spent-sorbent hazardous-waste fixating and cementitious properties

    SciTech Connect (OSTI)

    Schultz, T.D.; Berger, R.L.; Fishbein, K.

    1989-03-01

    The primary purpose of the project was to develop a use for the spent sorbent from a spray dryer flue gas desulfurization system. In addition to spent sorbent, fly ash was included in the utilization schemes because it is a byproduct of coal combustion and because it is a pozzolan. It would be helpful to find uses for these two substances and thus decrease the amount of land needed for their disposal and help offset the costs of flue gas desulfurization.

  12. Supplement Analysis … Spent Nuclear Fuel and SRS H-Canyon Operations

    Energy Savers [EERE]

    DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation U.S. Department of Energy Washington, DC November 2015 DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation 1.0 INTRODUCTION The Department of Energy (DOE) has a continuing responsibility for safeguarding

  13. DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Office of Environmental Management (EM)

    Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel October 2014 [This page left blank.] FOREWORD In January 2012, the final report of the Blue Ribbon Commission on America's Nuclear Future urged the Administration to conduct a review of the current policy to dispose of defense and commercial high level radioactive waste and spent nuclear fuel in a single repository or repositories. Based on that recommendation, in January 2013, the Administration's

  14. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    2000-10-12

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

  15. Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

    SciTech Connect (OSTI)

    Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

    2015-12-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  16. Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

    SciTech Connect (OSTI)

    Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

    2016-01-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  17. Annual report, FY 1979 Spent fuel and fuel pool component integrity.

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.

  18. Impacts of a high-burnup spent fuel on a geological disposal system design

    SciTech Connect (OSTI)

    Cho, D.K.; Lee, Y.; Lee, J.Y.; Choi, H.J.; Choi, J.W. [Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon-city (Korea, Republic of)

    2007-07-01

    The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill needed to constitute a disposal system were comprehensively analyzed based on the spent fuel inventory to generate 1 Terawatt-year (TWa), to establish the overall effects and consequences on a geological disposal. As a result, impact of a burnup increase on the criticality safety and radiation shielding was shown to be negligible. The disposal area, however, is considerably affected because of a higher thermal load. And, it is reasonable to use a canister such as the Korean Reference Disposal Canister (KDC-1) containing 4 spent fuels up to 50 GWD/MtU, and to use a canister containing 3 spent fuels beyond 50 GWD/MtU. Although a considerable increased, 33 % in the tunnel length and 30 % in the excavation volume, was observed as the burnup increases from 50 to 60 GWD/MtU, because a decrease in the canister needs can offset an increase in the excavation volume, it can be concluded that a burnup increase of a spent fuel is not a critical concern for a geological disposal of a spent fuel. (authors)

  19. Disposal of defense spent fuel and HLW at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1993-06-01

    Irradiated nuclear fuel has been reprocessed at the Idaho Chemical Processing Plant (ICPP) since 1953 to recover uranium-235 and krypton-85 for the US Department of Energy (DOE). The resulting acidic high-level radioactive waste (HLW) has been solidified to a calcine since 1963 and stored in stainless steel underground bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage at the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal.

  20. Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

    SciTech Connect (OSTI)

    Trond Bjornard; Philip C. Durst

    2012-05-01

    This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA) of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a countrys safeguards agreement with the IAEA. If these requirements are understood at the earliest stages of facility design, it will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards, and will help the IAEA implement nuclear safeguards worldwide, especially in countries building their first nuclear facilities. It is also hoped that this guidance document will promote discussion between the IAEA, State Regulator/SSAC, Project Design Team, and Facility Owner/Operator at an early stage to ensure that new ISFSIs will be effectively and efficiently safeguarded. This is intended to be a living document, since the international nuclear safeguards requirements may be subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and facility operators for greater efficiency and cost effectiveness. As these improvements are made, it is recommended that the subject guidance document be updated and revised accordingly.

  1. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect (OSTI)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the worlds first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  2. The EU Approach for Responsible and Safe Management of Spent Fuel and Radioactive Waste - 12118

    SciTech Connect (OSTI)

    Blohm-Hieber, Ute; Necheva, Christina [European Commission, Directorate-General for Energy, Luxembourg L-2920 (Luxembourg)

    2012-07-01

    In July 2011 legislation on responsible and safe management of spent fuel and radioactive waste was adopted in the European Union (EU). It aims at ensuring a high level of safety, avoiding undue burdens on future generations and enhancing transparency. EU Member States are responsible for the management of their spent fuel and/or radioactive waste. Each Member State remains free to define its fuel cycle policy. The spent fuel can be regarded either as a valuable resource that may be reprocessed or as radioactive waste that is destined for direct disposal. Whatever option is chosen, the disposal of high level waste, separated at reprocessing, or of spent fuel regarded as waste should be considered. The storage of radioactive waste, including long-term storage, is an interim solution, but not an alternative to disposal. To this end, each Member State has to establish, maintain and implement national policy, framework and programme for management of spent fuel and/or radioactive waste in the long term. Member States will invite international peer reviews to ensure that high safety standards are achieved. The EU approach is anchored in internationally endorsed principles and requirements of the IAEA safety standards and the Joint Convention and in this context makes them legally binding and enforceable in the EU. The EU approach of regulating the management of spent fuel and radioactive waste is anchored in the competence of the national regulatory authorities and in the internationally endorsed principles and requirements of the IAEA Safety Standards and the Joint Convention. Member States have to report to the Commission on the implementation of Directive 2011/70/Euratom for the first time by 23 August 2015, and every 3 years thereafter, taking advantage of the review and reporting under the Joint Convention. On the basis of the Member States' reports, the Commission will submit to the European Parliament and the Council a report on progress made and an inventory of radioactive waste and spent fuel present in the EU territory and the future prospects. Directive 2011/70/Euratom is a logical next step after the Council Directive 2009/71/Euratom on the nuclear safety of nuclear installations. The EU is the first major regional actor providing a binding legal framework on nuclear safety and on responsible and safe management of spent fuel and radioactive waste, and thus is a real model to progress spent fuel and waste management in a safe and responsible manner. (authors)

  3. Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany DOE/EA-1977 DRAFT ENVIRONMENTAL ASSESSMENT FOR THE ACCEPTANCE AND DISPOSITION OF SPENT NUCLEAR FUEL CONTAINING U.S.-ORIGIN HIGHLY ENRICHED URANIUM FROM THE FEDERAL REPUBLIC OF GERMANY January 2016 U.S. DEPARTMENT OF ENERGY SAVANNAH RIVER OPERATIONS OFFICE AIKEN, SOUTH CAROLINA Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing

  4. Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched

    Office of Environmental Management (EM)

    EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany DOE/EA-1977 DRAFT ENVIRONMENTAL ASSESSMENT FOR THE ACCEPTANCE AND DISPOSITION OF SPENT NUCLEAR FUEL CONTAINING U.S.-ORIGIN HIGHLY ENRICHED URANIUM FROM THE FEDERAL REPUBLIC OF GERMANY January 2016 U.S. DEPARTMENT OF ENERGY SAVANNAH RIVER OPERATIONS OFFICE AIKEN, SOUTH CAROLINA Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing

  5. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  6. Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2000-04-14

    The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

  7. Monte Carlo analysis of neutron slowing-down-time spectrometer for fast reactor spent fuel assay

    SciTech Connect (OSTI)

    Chen, Jianwei; Lineberry, Michael

    2007-07-01

    Using the neutron slowing-down-time method as a nondestructive assay tool to improve input material accountancy for fast reactor spent fuel reprocessing is under investigation at Idaho State University. Monte Carlo analyses were performed to simulate the neutron slowing down process in different slowing down spectrometers, namely, lead and graphite, and determine their main parameters. {sup 238}U threshold fission chamber response was simulated in the Monte Carlo model to represent the spent fuel assay signals, the signature (fission/time) signals of {sup 235}U, {sup 239}Pu, and {sup 241}Pu were simulated as a convolution of fission cross sections and neutron flux inside the spent fuel. {sup 238}U detector signals were analyzed using linear regression model based on the signatures of fissile materials in the spent fuel to determine weight fractions of fissile materials in the Advanced Burner Test Reactor spent fuel. The preliminary results show even though lead spectrometer showed a better assay performance than graphite, graphite spectrometer could accurately determine weight fractions of {sup 239}Pu and {sup 241}Pu given proper assay energy range were chosen. (authors)

  8. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

  9. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect (OSTI)

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as used nuclear fuel [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 Nm was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 Nm. Failure was observed in nine of the tested rod specimens. The cycles-to-failure ranged from 1.22 105 to 4.70 106, and the amplitudes varied from 15.2 to 7.6 Nm. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 Nm2. Online monitoring revealed that the flexural rigidity was slightly lower due to the higher loading, from 25 to 42 Nm2. Generally, no substantial change of rigidity was observed based on the online monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of sample lifetime with increasing amplitude is well defined.

  10. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM'S program specifications. (CBS)

  11. 105 K East and 105 K West fuel transfer bay crane use strategy for spent nuclear fuel path-forward

    SciTech Connect (OSTI)

    Ard, K.E.

    1996-04-02

    The purpose of this document is to outline the K Basins 30 ton crane qualification strategy for use in the Spent Nuclear Fuel Project fuel relocation campaign.

  12. Current State of Knowledge of Water Radiolysis Effects on Spent Nuclear Fuel Corrosion

    SciTech Connect (OSTI)

    Christensen, H.; Sunder, S.

    2000-07-15

    Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O{sub 2} or H{sub 2}O{sub 2} in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO{sub 2} in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study, it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed.

  13. A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL & TREATMENT & DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

    SciTech Connect (OSTI)

    PERES, M.W.

    2007-01-09

    Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel.

  14. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    SciTech Connect (OSTI)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  15. Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

    SciTech Connect (OSTI)

    Zhao, P.; Steward, S.A.

    1997-01-01

    Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations.

  16. RAIL ROUTING - CURRENT PRACTICES FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE SHIPMENTS,

    Office of Environmental Management (EM)

    4.31 Final 7/30 RAIL ROUTING - CURRENT PRACTICES FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE SHIPMENTS, AND A COMPARATIVE ANALYSIS OF HIGHWAY REGULATORY GUIDELINES APPLIED TO RAIL A DISCUSSION PAPER PREPARED BY THE RAIL TOPIC GROUP OF THE TRANSPORTATION EXTERNAL COORDINATION WORKING GROUP JULY 2004 1 Revision 4.31 Final 7/30 Rail Routing- Current Practices for Spent Nuclear Fuel And High-Level Waste Shipments, And a Comparative Analysis Of Highway Regulatory Guidelines Applied to Rail I.

  17. Spent fuel performance data: An analysis of data relevant to the NNWSI Project

    SciTech Connect (OSTI)

    Oversby, V.M.; Shaw, H.F.

    1987-08-01

    This paper summarizes the physical and chemical properties of spent light water reactor fuel that might influence its performance as a waste form under geologic disposal conditions at Yucca Mountain, Nevada. Results obtained on the dissolution testing of spent fuel conducted by the NNWSI Project are presented and discussed. Work published by other programs, in particular those of Canada and Sweden, are reviewed and compared with the NNWSI testing results. An attempt is made to relate all of the results to a common basis of presentation and to rationalize apparent conflicts between sets of results obtained under different experimental conditions.

  18. Extended Storage for Research and Test Reactor Spent Fuel for 2006 and Beyond

    SciTech Connect (OSTI)

    Hurt, William Lon; Moore, K.M.; Shaber, Eric Lee; Mizia, Ronald Eugene

    1999-10-01

    This paper will examine issues associated with extended storage of a variety of spent nuclear fuels. Recent experiences at the Idaho National Engineering and Environmental Laboratory and Hanford sites will be described. Particular attention will be given to storage of damaged or degraded fuel. The first section will address a survey of corrosion experience regarding wet storage of spent nuclear fuel. The second section will examine issues associated with movement from wet to dry storage. This paper also examines technology development needs to support storage and ultimate disposition.

  19. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    SciTech Connect (OSTI)

    Newsom, H.C.

    1999-01-24

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted.

  20. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Loiseau, O. (Institut de radioprotection et de Surete Nucleaire, France); Koch, W. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Billone, M. C. (Argonne National Laboratory, USA); Lucero, Daniel A.; Burtseva, T. (Argonne National Laboratory, USA); Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO{sub 2}, test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage--aerosol test program, performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission, had significant inputs from, and is strongly supported and coordinated by both the U.S. and international program participants in Germany, France, and the U.K., as part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC.

  1. Electrorefining Experience For Pyrochemical Reprocessing of Spent EBR-II Driver Fuel

    SciTech Connect (OSTI)

    S. X. Li; T. A. Johnson; B. R. Westphal; K. M. Goff; R. W. Benedict

    2005-10-01

    Pyrochemical processing has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor-II (EBR-II) at Idaho National Laboratory since 1996. This report summarizes technical advancements made in electrorefining of spent EBR-II driver fuel in the Mk-IV electrorefiner since the pyrochemical processing was integrated into the AFCI program in 2002. The significant advancements include improving uranium dissolution and noble metal retention from chopped fuel segments, increasing cathode current efficiency, and achieving co-collection of zirconium along with uranium from the cadmium pool.

  2. Assessment of the safety of spent fuel transportation in urban environs

    SciTech Connect (OSTI)

    Sandoval, R.P.; Weber, J.P.; Levine, H.S.; Romig, A.D.; Johnson, J.D.; Luna, R.E.; Newton, G.J.; Wong, B.A.; Marshall, R.W. Jr.; Alvarez, J.L.

    1983-06-01

    The results of a program to provide an experimental data base for estimating the radiological consequences from a hypothetical sabotage attack on a light-water-reactor spent fuel shipping cask in a densely populated area are presented. The results of subscale and full-scale experiments in conjunction with an analytical modeling study are described. The experimental data were used as input to a reactor-safety consequence model to predict radiological health consequences resulting from a hypothetical sabotage attack on a spent-fuel shipping cask in the Manhattan borough of New York City. The results of these calculations are presented.

  3. Basis for Identification of Disposal Options for R and D for Spent Nuclear

    Energy Savers [EERE]

    Fuel and High-Level Waste | Department of Energy Basis for Identification of Disposal Options for R and D for Spent Nuclear Fuel and High-Level Waste Basis for Identification of Disposal Options for R and D for Spent Nuclear Fuel and High-Level Waste The Used Fuel Disposition campaign (UFD) is selecting a set of geologic media for further study that spans a suite of behavior characteristics that impose a broad range of potential conditions on the design of the repository, the engineered

  4. EM Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin

    Office of Environmental Management (EM)

    at Idaho Site | Department of Energy Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site EM Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site August 31, 2015 - 12:10pm Addthis Workers train to prepare for the ion-exchange water treatment process. Workers train to prepare for the ion-exchange water treatment process. The ion-exchange resin used to treat the storage pool water. The ion-exchange resin used to treat the

  5. Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

    SciTech Connect (OSTI)

    Patrick, W.C.

    1986-03-30

    In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.

  6. DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2.

  7. Imaging Spent Fuel in Dry Storage Casks with Cosmic Ray Muons

    SciTech Connect (OSTI)

    Durham, J. Matthew; Dougan, Arden

    2015-11-05

    Highly energetic cosmic ray muons are a natural source of ionizing radiation that can be used to make tomographic images of the interior of dense objects. Muons are capable of penetrating large amounts of shielding that defeats typical radiographic probes like neutrons or photons. This is the only technique which can examine spent nuclear fuel rods sealed inside dry casks.

  8. FIELD-DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-09-12

    Methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of aqueous spent fuel storage basins and determine the oxide thickness on the spent fuel basin materials were developed to assess the corrosion potential of a basin. this assessment can then be used to determine the amount of time fuel has spent in a storage basin to ascertain if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations and assist in evaluating general storage basin operations. The test kit was developed based on the identification of key physical, chemical and microbiological parameters identified using a review of the scientific and basin operations literature. The parameters were used to design bench scale test cells for additional corrosion analyses, and then tools were purchased to analyze the key parameters. The tools were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The sampling kit consisted of a total organic carbon analyzer, an YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization confirmed that the L Area basin is a well operated facility with low corrosion potential.

  9. Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

    SciTech Connect (OSTI)

    Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy; Schmieman, Eric

    2013-07-01

    Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

  10. An overview of spent-fuel processing in the global nuclear-energy partnership

    SciTech Connect (OSTI)

    Laidler, James J.

    2008-07-01

    Spent nuclear fuel is being generated at a prodigious rate in the U.S. and in other countries with robust nuclear-power-generation infrastructures, and the annual rate of production is likely to triple by 2050. The U.S. is engaged in the development of commercial light-water-reactor spent- fuel-treatment processes that are intended to meet certain rigorous criteria for separations efficiency, waste management benefits, and economy of industrial-scale operations. Aqueous solvent-extraction processes are the technology of choice, and a variety of process options have been designed and tested for technical feasibility. In general, the processes involve substantial partitioning of the constituents of spent nuclear fuel, so that effective use can be made of the recovered unburned uranium and other fissile isotopes that can be recycled as fuel for contemporary or advanced reactors. Those constituents that are destined for disposal as waste are also separated in order that they can be placed into durable waste forms that are expressly tailored for a particular disposition pathway. The U.S. is also working with international partners as part of the Global Nuclear Energy Partnership (GNEP) to develop a consistent worldwide approach to the treatment of spent fuel and the disposition of wastes arising from such processing. (authors)

  11. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

  12. Multi-Canister overpack pressurization monitoring and control methodology for the spent nuclear fuel project

    SciTech Connect (OSTI)

    Pajunen, A.L., Westinghouse Hanford

    1996-07-19

    A control methodology is developed and monitoring alternatives evaluated for controlling pressurization in a Multi- Canister Overpack for the Hanford Spent Nuclear Fuel Project. Monitoring alternative evaluations include concept description, identification of uncertainties, and identification of experimental work required for implementation. A monitoring alternative is recommended and implementation requirements, risks and start up testing associated with the recommendation are discussed.

  13. LAB STUDY ON REGENERATION OF SPENT DOWEX 21K 16-20 MESH ION EXCHANGE RESIN

    SciTech Connect (OSTI)

    DUNCAN, J.B.

    2007-01-24

    Currently the effort to remove chromate from groundwater in the 100K and 100H Areas uses DOWEX 21K 16-20. This report addresses the procedure and results of a laboratory study for regeneration of the spent resin by sodium hydroxide, sulfuric acid, or sodium sulfate to determine if onsite regeneration by the Effluent Treatment Facility is a feasible option.

  14. Characterization program management plan for Hanford K basin spent nuclear fuel

    SciTech Connect (OSTI)

    TRIMBLE, D.J.

    1999-07-19

    The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng.

  15. Characterization program management plan for Hanford K Basin spent nuclear fuel

    SciTech Connect (OSTI)

    Lawrence, L.A.

    1998-05-14

    The management plan developed to characterize the K Basin Spent Nuclear Fuel was revised to incorporate actions necessary to comply with the Office of Civilian Radioactive Waste Management Quality Assurance Requirements Document 0333P. This plan was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. This revision to the Program Management Plan replaces Westinghouse Hanford Company with Duke Engineering and Services Hanford, Inc., updates the various activities where necessary, and expands the Quality Assurance requirements to meet the applicable requirements document. Characterization will continue to utilize the expertise and capabilities of both organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for Duke Engineering and Services Hanford, Inc. and Pacific Northwest National Laboratory to support the Spent Nuclear Fuels Project at Hanford.

  16. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  17. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

  18. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  19. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  20. Method of separating and recovering uranium and related cations from spent Purex-type systems

    DOE Patents [OSTI]

    Mailen, J.C.; Tallent, O.K.

    1987-02-25

    A process for separating uranium and related cations from a spent Purex-type solvent extraction system which contains degradation complexes of tributylphosphate wherein the system is subjected to an ion-exchange process prior to a sodium carbonate scrubbing step. A further embodiment comprises recovery of the separated uranium and related cations. 5 figs.

  1. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  2. What to Expect When Readying to Move Spent Nuclear Fuel from Commercial

    Broader source: Energy.gov (indexed) [DOE]

    Nuclear Power Plants | Department of Energy PDF icon What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project Overview Indiana Department of Homeland Security - NNPP Exercise Better Security Through Discussion

  3. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  4. A Second Look at Neutron Resonance Transmission Analysis as a Spent Fuel NDA Technique

    SciTech Connect (OSTI)

    James W .Sterbentz; David L. Chichester

    2011-07-01

    Many different nondestructive analysis techniques are currently being investigated as a part of the United States Department of Energy's Next Generation Safeguards Initiative (NGSI) seeking methods to quantify plutonium in spent fuel. Neutron Resonance Transmission Analysis (NRTA) is one of these techniques. Having first been explored in the mid-1970s for the analysis of individual spent-fuel pins a second look, using advanced simulation and modeling methods, is now underway to investigate the suitability of the NRTA technique for assaying complete spent nuclear fuel assemblies. The technique is similar to neutron time-of-flight methods used for cross-section determinations but operates over only the narrow 0.1-20 eV range where strong, distinguishable resonances exist for both the plutonium (239, 240, 241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Initial modeling shows excellent agreement with previously published experimental data for measurements of individual spent-fuel pins where plutonium assays were demonstrated to have a precision of 2-4%. Within the simulation and modeling analyses of this project scoping studies have explored fourteen different aspects of the technique including the neutron source, drift tube configurations, and gross neutron transmission as well as the impacts of fuel burn up, cooling time, and fission-product interferences. These results show that NRTA may be a very capable experimental technique for spent-fuel assay measurements. The results suggest sufficient transmission strength and signal differentiability is possible for assays through up to 8 pins. For an 8-pin assay (looking at an assembly diagonally), 64% of the pins in a typical 17 ? 17 array of a pressurized water reactor fuel assembly can be part of a complete transmission assay measurement with high precision. Analysis of rows with up to 12 pins may also be feasible but with diminished precision. Preliminary data analysis of an NRTA simulation has demonstrated the simplicity of the technique.

  5. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and coordinated by both the U.S. and international program participants in Germany, France, and others, as part of the International Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC).

  6. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    SciTech Connect (OSTI)

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

  7. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    SciTech Connect (OSTI)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)

  8. Further Evaluation of the Neutron Resonance Transmission Analysis (NRTA) Technique for Assaying Plutonium in Spent Fuel

    SciTech Connect (OSTI)

    J. W. Sterbentz; D. L. Chichester

    2011-09-01

    This is an end-of-year report (Fiscal Year (FY) 2011) for the second year of effort on a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The second-year goals for this project included: (1) assessing the neutron source strength needed for the NRTA technique, (2) estimating count times, (3) assessing the effect of temperature on the transmitted signal, (4) estimating plutonium content in a spent fuel assembly, (5) providing a preliminary assessment of the neutron detectors, and (6) documenting this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes work performed over a nine month period from January-September 2011 and is to be considered a follow-on or add-on report to our previous published summary report from December 2010 (INL/EXT-10-20620).

  9. Spent fuel dissolution rates as a function of burnup and water chemistry

    SciTech Connect (OSTI)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  10. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    SciTech Connect (OSTI)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  11. Assessment of spent-fuel waste-form/stabilizer alternatives for geologic disposal

    SciTech Connect (OSTI)

    Einziger, R.E.; Himes, D.A.

    1982-06-01

    The Office of Nuclear Waste Isolation (ONWI) is studying the possibility of burying canisterized unreprocessed spent fuel in a deep geologic repository. One aspect of this study is an assessment of the possible spent fuel waste forms. The fuel performance portion of the Waste Form Assessment was to evaluate five candidate spent fuel waste forms for postemplacement performance with emphasis on their ability to retard the release of radionuclides to the repository geology. Spent fuel waste forms under general consideration were: (1) unaltered fuel assembly; (2) fuel assembly with end fittings removed to shorten the length; (3 rods vented to remove gases and resealed; (4) disassembled fuel bundles to close-pack the rods; and (5) rods chopped and fragments immobilized in a matrix material. Thirteen spent fuel waste forms, classified by generic stabilizer type, were analyzed for relative in-repository performance based on: (1) waste form/stabilizer support against lithostatic pressure; (2) long-term stability for radionuclide retention; (3) minimization of cladding degradation; (4) prevention of canister/repository breach due to pressurization; (5) stabilizer heat transfer; (6) the stabilizer as an independent barrier to radionuclide migration; and (7) prevention of criticality. The waste form candidates were ranked as follows: (1) the best waste form/stabilizer combination is the intact assembly, with or without end bells, vented (and resealed) or unvented, with a solid stabilizer; (2) a suitable alternative is the combination of bundled close-packed rods with a solid stabilizer around the outside of the bundle to resist lithostatic pressure; and (3) the other possible waste forms are of lower ranking with the worst waste form/stabilizer combination being the intact assembly with a gas stabilizer or the chopped fuel.

  12. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  13. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  14. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    SciTech Connect (OSTI)

    Dilger, Fred; Halstead, Robert J.; Ballard, James D.

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

  15. Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

    SciTech Connect (OSTI)

    Reimus, P.W.; Simonson, S.A.

    1988-04-01

    This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B.

  16. Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

    DOE Patents [OSTI]

    Heckman, R.A.

    1980-12-19

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  17. Recovery and regeneration of spent MHD seed material by the formate process

    DOE Patents [OSTI]

    Sheth, A.C.; Holt, J.K.; Rasnake, D.G.; Solomon, R.L.; Wilson, G.L.; Herrigel, H.R.

    1991-10-15

    The specification discloses a spent seed recovery and regeneration process for an MHD power plant employing an alkali metal salt seed material such as potassium salt wherein the spent potassium seed in the form of potassium sulfate is collected from the flue gas and reacted with calcium hydroxide and carbon monoxide in an aqueous solution to cause the formation of calcium sulfate and potassium formate. The pH of the solution is adjusted to suppress formation of formic acid and to promote precipitation of any dissolved calcium salts. The solution containing potassium formate is then employed to provide the potassium salt in the form of potassium formate or, optionally, by heating the potassium formate under oxidizing conditions to convert the potassium formate to potassium carbonate. 5 figures.

  18. Recovery and regeneration of spent MHD seed material by the formate process

    DOE Patents [OSTI]

    Sheth, Atul C. (Tullahoma, TN); Holt, Jeffrey K. (Manchester, TN); Rasnake, Darryll G. (Manchester, TN); Solomon, Robert L. (Seattle, WA); Wilson, Gregory L. (Redmond, WA); Herrigel, Howard R. (Seattle, WA)

    1991-01-01

    The specification discloses a spent seed recovery and regeneration process for an MHM power plant employing an alkali metal salt seed material such as potassium salt wherein the spent potassium seed in the form of potassium sulfate is collected from the flue gas and reacted with calcium hydroxide and carbon monoxide in an aqueous solution to cause the formation of calcium sulfate and potassium formate. The pH of the solution is adjusted to supress formation of formic acid and to promote precipitation of any dissolved calcium salts. The solution containing potassium formate is then employed to provide the potassium salt in the form of potassium formate or, optionally, by heating the potassium formate under oxidizing conditions to convert the potassium formate to potassium carbonate.

  19. A metallic fuel cycle concept from spent oxide fuel to metallic fuel

    SciTech Connect (OSTI)

    Fujita, Reiko; Kawashima, Masatoshi; Yamaoka, Mitsuaki; Arie, Kazuo; Koyama, Tadafumi

    2007-07-01

    A Metallic fuel cycle concept for Self-Consistent Nuclear Energy System (SCNES) has been proposed in a companion papers. The ultimate goal of the SCNES is to realize sustainable energy supply without endangering the environment and humans. For future transition period from LWR era to SCNES era, a new metallic fuel recycle concept from LWR spent fuel has been proposed in this paper. Combining the technology for electro-reduction of oxide fuels and zirconium recovery by electrorefining in molten salts in the nuclear recycling schemes, the amount of radioactive waste reduced in a proposed metallic fuel cycle concept. If the recovery ratio of zirconium metal from the spent zirconium waste is 95%, the cost estimation in zirconium recycle to the metallic fuel materials has been estimated to be less than 1/25. (authors)

  20. Geologic repository design and disposal: GNEP spent fuel processing-waste volume

    SciTech Connect (OSTI)

    Bauer, T.H.; Wigeland, R.A.

    2007-07-01

    Previous work has shown that removal of key heat generating elements from spent fuel would allow greater utilization of space in a geologic repository such as Yucca Mountain by factors of 100 or more without increasing the estimated peak dose rate to an exposed individual. However, achieving such utilization increases within a repository storage drift requires the density of the remaining fission products, actinide elements, etc. to be increased by roughly the same factor as the utilization increase, itself. This paper analyzes several alternative drift configurations possible within a designated repository area that could: (1) allow greater volume for waste storage and (2) maintain significant utilization benefit. For a representative range of GNEP-generated waste streams, computed results show that increase in repository area space utilization by a factor {approx}100 can be maintained with such configurations as long as waste stream volume can be reduced from that of the original spent fuel by a factor of {approx}10. (authors)

  1. Advanced spent fuel conditioning process (ACP) progress with respect to remote operation and maintenance

    SciTech Connect (OSTI)

    Lee, Hyo Jik; Lee, Jong Kwang; Park, Byung Suk; Yoon, Ji Sup

    2007-07-01

    Korea Atomic Energy Research Institute (KAERI) has been developing an Advanced Spent Fuel Conditioning Process (ACP) to reduce the volume of spent fuel, and the construction of the ACP facility (ACPF) for a demonstration of its technical feasibility has been completed. In 2006 two inactive demonstrations were performed with simulated fuels in the ACPF. Accompanied by process equipment performance tests, its remote operability and maintainability were also tested during that time. Procedures for remote operation tasks are well addressed in this study and evaluated thoroughly. Also, remote maintenance and repair tasks are addressed regarding some important modules with a high priority order. The above remote handling test's results provided a lot of information such as items to be revised to improve the efficiency of the remote handling tasks. This paper deals with the current status of ACP and the progress of remote handling of ACPF. (authors)

  2. Non-Destructive Spent Fuel Characterization with Semi-Conducting Gallium Arsinde Neutron Imaging Arrays

    SciTech Connect (OSTI)

    Douglas S. McGregor; Holly K. Gersch; Jeffrey D. Sanders; John C. Lee; Mark D. Hammig; Michael R. Hartman; Yong Hong Yang; Raymond T. Klann; Brian Van Der Elzen; John T. Lindsay; Philip A. Simpson

    2002-01-30

    High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency.

  3. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  4. Method for processing aluminum spent potliner in a graphite electrode ARC furnace

    DOE Patents [OSTI]

    O'Connor, William K. (Lebanon, OR); Turner, Paul C. (Independence, OR); Addison, Gerald W. (St. Stephen, SC)

    2002-12-24

    A method of processing spent aluminum pot liner containing carbon, cyanide compositions, fluorides and inorganic oxides. The spent aluminum pot liner is crushed iron oxide is added to form an agglomerated material. The agglomerated material is melted in an electric arc furnace having the electrodes submerged in the molten material to provide a reducing environment during the furnace operation. In the reducing environment, pot liner is oxidized while the iron oxides are reduced to produce iron and a slag substantially free of cyanide compositions and fluorides. An off-gas including carbon oxides and fluorine is treated in an air pollution control system with an afterburner and a scrubber to produce NaF, water and a gas vented to the atmosphere free of cyanide compositions, fluorine and CO.

  5. Retorting of oil shale followed by solvent extraction of spent shale: Experiment and kinetic analysis

    SciTech Connect (OSTI)

    Khraisha, Y.H.

    2000-05-01

    Samples of El-Lajjun oil shale were thermally decomposed in a laboratory retort system under a slow heating rate (0.07 K/s) up to a maximum temperature of 698--773 K. After decomposition, 0.02 kg of spent shale was extracted by chloroform in a Soxhlet extraction unit for 2 h to investigate the ultimate amount of shale oil that could be produced. The retorting results indicate an increase in the oil yields from 3.24% to 9.77% of oil shale feed with retorting temperature, while the extraction results show a decrease in oil yields from 8.10% to 3.32% of spent shale. The analysis of the data according to the global first-order model for isothermal and nonisothermal conditions shows kinetic parameters close to those reported in literature.

  6. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.; Coops, M.S.

    1982-01-19

    A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

  7. Plan for characterization of K Basin spent nuclear fuel and sludge

    SciTech Connect (OSTI)

    Lawrence, L.A.; Marschman, S.C.

    1995-06-01

    This plan outlines a characterization program that supports the accelerated Path Forward scope and schedules for the Spent Nuclear Fuel stored in the Hanford K Basins. This plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years and is limited to in-situ and laboratory examinations of the spent nuclear fuel and sludge in the K East and K West Basins. The program provides bounding behavior of the fuel, and verification and acceptability for three different sludge disposal pathways. Fuel examinations are based on two shipping campaigns for the K West Basin and one from the K East Basin. Laboratory examinations include physical condition, hydride and oxide content, conditioning testing, and dry storage behavior.

  8. Microstructural characteristics of PWR spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1985-11-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  9. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  10. A method for determining the spent-fuel contribution to transport cask containment requirements

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Duffey, T.A.; Sutherland, S.H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  11. Long-term kinetic effects and colloid formations in dissolution of LWR spent fuels

    SciTech Connect (OSTI)

    Ahn, T.M.

    1996-11-01

    This report evaluates continuous dissolution and colloid formation during spent-fuel performance under repository conditions in high-level waste disposal. Various observations suggest that reprecipitated layers formed on spent-fuel surfaces may not be protective. This situation may lead to continuous dissolution of highly soluble radionuclides such as C-14, Cl-36, Tc-99, I-129, and Cs-135. However, the diffusion limits of various species involved may retard dissolution significantly. For low-solubility actinides such as Pu-(239+240) or Am-(241+243), various processes regarding colloid formation have been analyzed. The processes analyzed are condensation, dispersion, and sorption. Colloid formation may lead to significant releases of low-solubility actinides. However, because there are only limited data available on matrix dissolution, colloid formation, and solubility limits, many uncertainties still exist. These uncertainties must be addressed before the significance of radionuclide releases can be determined. 118 refs.

  12. Determining plutonium mass in spent fuel using Cf-252 interrogation with prompt neutron detection

    SciTech Connect (OSTI)

    Hu, Jianwei; Tobin, Stephen J; Menlove, Howard O; Croft, Stephen

    2010-01-01

    {sup 252}Cf Interrogation with Prompt Neutron (CIPN) detection is proposed as one of 14 NDA techniques to determine Pu mass in spent fuel assemblies (FAs). CIPN is a low-cost and portable instrument, and it looks like a modified fork detector combined with an active interrogation source. Fission chamber (FC) is chosen as neutron detector because of its insensitivity to {gamma} radiation. The CIPN assay is comprised of two measurements, a background count and an active count, without and with the {sup 252}Cf source next to the fuel respectively. The net signal above background is primarily due to the multiplication of Cf source neutrons caused by the fissile content. The capability of CIPN to detect diversion and to determine fissile content was quantified using MCNPX simulations. New schemes were proposed (such as burnup and cooling time correction, etc.) and the results show that the fissile content of a target spent fuel assembly can be determined using CIPN signal.

  13. Chemical reactivity testing for the National Spent Nuclear Fuel Program. Revision 2

    SciTech Connect (OSTI)

    Koester, L.W.

    2000-02-08

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, Y60-101PD, Quality Program Description, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted. The project consists of conducting three separate series of related experiments, ''Passivation of Uranium Hydride Powder With Oxygen and Water'', '''Passivation of Uranium Hydride Powder with Surface Characterization'', and ''Electrochemical Measure of Uranium Hydride Corrosion Rate''.

  14. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Pruett, D.J.; McTaggart, D.R.

    1983-08-31

    Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

  15. Electrochemical performance studies of MnO{sub 2} nanoflowers recovered from spent battery

    SciTech Connect (OSTI)

    Ali, Gomaa A.M.; Tan, Ling Ling; Jose, Rajan; Yusoff, Mashitah M.; Chong, Kwok Feng

    2014-12-15

    Highlights: MnO{sub 2} is recovered from spent zinccarbon batteries as nanoflowers structure. Recovered MnO{sub 2} nanoflowers show high specific capacitance. Recovered MnO{sub 2} nanoflowers show stable electrochemical cycling up to 900 cycles. Recovered MnO{sub 2} nanoflowers show low resistance in EIS data. - Abstract: The electrochemical performance of MnO{sub 2} nanoflowers recovered from spent household zinccarbon battery is studied by cyclic voltammetry, galvanostatic charge/discharge cycling and electrochemical impedance spectroscopy. MnO{sub 2} nanoflowers are recovered from spent zinccarbon battery by combination of solution leaching and electrowinning techniques. In an effort to utilize recovered MnO{sub 2} nanoflowers as energy storage supercapacitor, it is crucial to understand their structure and electrochemical performance. X-ray diffraction analysis confirms the recovery of MnO{sub 2} in birnessite phase, while electron microscopy analysis shows the MnO{sub 2} is recovered as 3D nanostructure with nanoflower morphology. The recovered MnO{sub 2} nanoflowers exhibit high specific capacitance (294 F g{sup ?1} at 10 mV s{sup ?1}; 208.5 F g{sup ?1} at 0.1 A g{sup ?1}) in 1 M Na{sub 2}SO{sub 4} electrolyte, with stable electrochemical cycling. Electrochemical data analysis reveal the great potential of MnO{sub 2} nanoflowers recovered from spent zinccarbon battery in the development of high performance energy storage supercapacitor system.

  16. A new paradigm: near-complete recycling of spent fuel - A path to sustainable nuclear energy

    SciTech Connect (OSTI)

    Del Cul, Guillermo D.; Spencer, Barry B.; Collins, Emory D.

    2007-07-01

    Recent studies indicate that maximized recycling, where more than 95% of the components of spent nuclear fuel are reused, can be economically justified and can reduce the mass of waste products by a substantial amount. The potentially removable and reusable components include the uranium, zirconium from the cladding, structural hardware, certain noble metal fission products, and the transuranic radionuclides. The approach to maximizing recycle and minimizing emissions and wastes should improve public acceptance of nuclear energy. (authors)

  17. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect (OSTI)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  18. Brittle Failure Design Criteria for Ductile Cast Iron Spent-Fuel

    Office of Scientific and Technical Information (OSTI)

    Recommendations for Ductile and Brittle Failure Design Criteria for Ductile Cast Iron Spent-Fuel Shipping Containers This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsi- bility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process

  19. Fourth National Report for the Joint Convention on the Safety of Spent Fuel

    Broader source: Energy.gov (indexed) [DOE]

    Management and on the Safety of Radioactive Waste Management | Department of Energy This Fourth United States of America (U.S.) National Report updates the Third Report published in October 2008, under the terms of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention). This report reflects developments in the U.S. through June 2011. This report satisfies the requirements of the Joint Convention for reporting on the

  20. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Pruett, David J. (Knoxville, TN); McTaggart, Donald R. (Knoxville, TN)

    1984-01-01

    Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

  1. The CASTOR-V/21 PWR spent-fuel storage cask: Testing and analyses: Interim report

    SciTech Connect (OSTI)

    Dziadosz, D.; Moore, E.V.; Creer, J.M.; McCann, R.A.; McKinnon, M.A.; Tanner, J.E.; Gilbert, E.R.; Goodman, R.L.; Schoonen, D.H.; Jensen, M.

    1986-11-01

    A performance test of a Gesellschaft fuer Nuklear Service CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask was performed. The test was the first of a series of cask performance tests planned under a cooperative agreement between Virginia Power and the US Department of Energy. The performance test consisted of loading the CASTOR-V/21 cask with 21 PWR spent fuel assemblies from Virginia Power's Surry reactor. Cask surface and fuel assembly guide tube temperatures, and cask surface gamma and neutron dose rates were measured. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Limited spent fuel integrity data were also obtained. Results of the performance test indicate the CASTOR-V/21 cask exhibited exceptionally good heat transfer performance which exceeded design expectations. Peak cladding temperatures with helium and nitrogen backfills in a vertical cast orientation and with helium in a horizontal orientation were less than the allowable of 380/sup 0/C with a total cask heat load of 28 kW. Significant convection heat transfer was present in vertical nitrogen and helium test runs as indicated by peak temperatures occurring in the upper regions of the fuel assemblies. Pretest temperature predictions of the HYDRA heat transfer computer program were in good agreement with test data, and post-test predictions agreed exceptionally well (25/sup 0/C) with data. Cask surface gamma and neutron dose rates were measured to be less than the design goal of 200 mrem/h. Localized peaks as high as 163 mrem/h were measured on the side of the cask, but peak dose rates of <75 mrem/h can easily be achieved with minor refinements to the gamma shielding design. From both heat transfer and shielding perspectives, the CASTOR-V/21 cask can, with minor refinements, be effectively implemented at reactor sites and central storage facilities for safe storage of spent fuel.

  2. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    SciTech Connect (OSTI)

    Linge, I. I.; Mitenkova, E. F. Novikov, N. V.

    2012-12-15

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  3. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  4. A GAMMA RAY SCANNING APPROACH TO QUANTIFY SPENT FUEL CASK RADIONUCLIDE CONTENTS

    SciTech Connect (OSTI)

    Branney, S.

    2011-07-01

    The International Atomic Energy Agency (IAEA) has outlined a need to develop methods of allowing re-verification of LWR spent fuel stored in dry storage casks without the need of a reference baseline measurement. Some scanning methods have been developed, but improvements can be made to readily provide required data for spent fuel cask verification. The scanning process should be conditioned to both confirm the contents and detect any changes due to container/contents degradation or unauthorized removal or tampering. Savannah River National Laboratory and The University of Tennessee are exploring a new method of engineering a high efficiency, cost effective detection system, capable of meeting the above defined requirements in a variety of environmental situations. An array of NaI(Tl) detectors, arranged to form a 'line scan' along with a matching array of 'honeycomb' collimators provide a precisely defined field of view with minimal degradation of intrinsic detection efficiency and with significant scatter rejection. Scanning methods are adapted to net optimum detection efficiency of the combined system. In this work, and with differing detectors, a series of experimental demonstrations are performed that map system spatial performance and counting capability before actual spent fuel cask scans are performed. The data are evaluated to demonstrate the prompt ability to identify missing fuel rods or other content abnormalities. To also record and assess cask tampering, the cask is externally examined utilizing FTIR hyper spectral and other imaging/sensing approaches. This provides dated records and indications of external abnormalities (surface deposits, smears, contaminants, corrosion) attributable to normal degradation or to tampering. This paper will describe the actual gathering of data in both an experimental climate and from an actual spent fuel dry storage cask, and how an evaluation may be performed by an IAEA facility inspector attempting to draw an independent safeguards conclusion concerning the status of the special nuclear material.

  5. Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Orland Park, IL); Miller, William E. (Naperville, IL); Wolson, Raymond D. (Lockport, IL); Gay, Eddie C. (Park Forest, IL)

    1991-01-01

    An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

  6. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    SciTech Connect (OSTI)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  7. Disposal options for burner ash from spent graphite fuel. Final study report November 1993

    SciTech Connect (OSTI)

    Pinto, A.P.

    1994-08-01

    Three major disposal alternatives are being considered for Fort St. Vrain Reactor (FSVR) and Peach Bottom Reactor (PBR) spent fuels: direct disposal of packaged, intact spent fuel elements; (2) removal of compacts to separate fuel into high-level waste (HLW) and low-level waste (LLW); and (3) physical/chemical processing to reduce waste volumes and produce stable waste forms. For the third alternative, combustion of fuel matrix graphite and fuel particle carbon coatings is a preferred technique for head-end processing as well as for volume reduction and chemical pretreatment prior to final fixation, packaging, and disposal of radioactive residuals (fissile and fertile materials together with fission and activation products) in a final repository. This report presents the results of a scoping study of alternate means for processing and/or disposal of fissile-bearing particles and ash remaining after combustion of FSVR and PBR spent graphite fuels. Candidate spent fuel ash (SFA) waste forms in decreasing order of estimated technical feasibility include glass-ceramics (GCs), polycrystalline ceramic assemblages (PCAs), and homogeneous amorphous glass. Candidate SFA waste form production processes in increasing order of estimated effort and cost for implementation are: low-density GCs via fuel grinding and simultaneous combustion and waste form production in a slagging cyclone combustor (SCC); glass or low-density GCs via fluidized bed SFA production followed by conventional melting of SFA and frit; PCAs via fluidized bed SFA production followed by hot isostatic pressing (HIPing) of SFA/frit mixtures; and high-density GCs via fluidized bed SFA production followed by HIPing of Calcine/Frit/SFA mixtures.

  8. American National Standard: design criteria for an independent spent-fuel-storage installation (water pool type)

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    This standard provides design criteria for systems and equipment of a facility for the receipt and storage of spent fuel from light water reactors. It contains requirements for the design of major buildings and structures including the shipping cask unloading and spent fuel storage pools, cask decontamination, unloading and loading areas, and the surrounding buildings which contain radwaste treatment, heating, ventilation and air conditioning, and other auxiliary systems. It contains requirements and recommendations for spent fuel storage racks, special equipment and area layout configurations, the pool structure and its integrity, pool water cleanup, ventilation, residual heat removal, radiation monitoring, fuel handling equipment, cask handling equipment, prevention of criticality, radwaste control and monitoring systems, quality assurance requirements, materials accountability, and physical security. Such an installation may be independent of both a nuclear power station and a reprocessing facility or located adjacent to any of these facilities in order to share selected support systems. Support systems shall not include a direct means of transferring fuel assemblies from the nuclear facility to the installation.

  9. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    SciTech Connect (OSTI)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems.

  10. Removal of arsenic, vanadium and/or nickel compounds from spent catecholated polymer

    DOE Patents [OSTI]

    Fish, R.H.

    1987-04-21

    Described is a process for removing arsenic, vanadium, and/or nickel from petroliferous derived liquids by contacting said liquid at an elevated temperature with a divinylbenzene-crosslinked polystyrene having catechol ligands anchored thereon. For vanadium and nickel removal an amine, preferably a diamine is included. Also, described is a process for regenerating spent catecholated polystyrene by removal of the arsenic, vanadium, and/or nickel bound to it from contacting petroliferous liquid as described above and involves: treating the spent polymer containing any vanadium and/or nickel with an aqueous acid to achieve an acid pH; and, separating the solids from the liquid; and then treating said spent catecholated polystyrene, at a temperature in the range of about 20 to 100 C with an aqueous solution of at least one carbonate and/or bicarbonate of ammonium, alkali and alkaline earth metals, said solution having a pH between about 8 and 10; and, separating the solids and liquids from each other. Preferably the regeneration treatment of arsenic containing catecholated polymer is in two steps wherein the first step is carried out with an aqueous alcoholic carbonate solution containing lower alkyl alcohol, and, the steps are repeated using a bicarbonate.

  11. Removal of arsenic, vanadium and/or nickel compounds from spent catecholated polymer

    DOE Patents [OSTI]

    Fish, Richard H.

    1987-01-01

    Described is a process for removing arsenic, vanadium, and/or nickel from petroliferous derived liquids by contacting said liquid at an elevated temperature with a divinylbenzene-crosslinked polystyrene having catechol ligands anchored thereon. For vanadium and nickel removal an amine, preferably a diamine is included. Also, described is a process for regenerating spent catecholated polystyrene by removal of the arsenic, vanadium, and/or nickel bound to it from contacting petroliferous liquid as described above and involves: treating the spent polymer containing any vanadium and/or nickel with an aqueous acid to achieve an acid pH; and, separating the solids from the liquid; and then treating said spent catecholated polystyrene, at a temperature in the range of about 20.degree. to 100.degree. C. with an aqueous solution of at least one carbonate and/or bicarbonate of ammonium, alkali and alkaline earth metals, said solution having a pH between about 8 and 10; and, separating the solids and liquids from each other. Preferably the regeneration treatment of arsenic containing catecholated polymer is in two steps wherein the first step is carried out with an aqueous alcoholic carbonate solution containing lower alkyl alcohol, and, the steps are repeated using a bicarbonate.

  12. Response of a Spent Fuel Transportation Cask to a Tunnel Fire Event

    SciTech Connect (OSTI)

    Bajwa, C. S.

    2003-02-25

    The staff of the Spent Fuel Project Office at the U.S. Nuclear Regulatory Commission undertook the investigation and thermal analysis of the Baltimore tunnel fire event. This event occurred in the Howard Street tunnel, in Baltimore, Maryland, on July 18, 2001. The staff was tasked with assessing the consequences of this event on the transportation of spent nuclear fuel. This paper describes the staff's coordination with the following government and laboratory organizations: the National Transportation Safety Board (NTSB), to determine the details of the train derailment and fire; the National Institute of Standards and Technology (NIST), to quantify the thermal conditions within the tunnel; the Center for Nuclear Waste Regulatory Analysis (CNWRA), to validate the NIST evaluations, and the Pacific Northwest National Laboratory (PNNL), to assist in the thermal analysis. The results of the staff's review and analysis efforts are also discussed. The staff has concluded that had the spent fuel transportation cask analyzed, a design approved under 10 CFR Part 71, been subjected to the Howard Street tunnel fire, no release of radioactive materials would have resulted from this postulated event, and the health and safety of the public would have been maintained.

  13. Engineering study for the treatment of spent ion exchange resin resulting from nuclear process applications

    SciTech Connect (OSTI)

    Place, B.G.

    1990-09-01

    This document is an engineering study of spent ion exchange resin treatment processes with the purpose of identifying one or more suitable treatment technologies. Classifications of waste considered include all classes of low-level waste (LLW), mixed LLW, transuranic (TRU) waste, and mixed TRU waste. A total of 29 process alternatives have been evaluated. Evaluation parameters have included economic parameters (both total life-cycle costs and capital costs), demonstrated operability, environmental permitting, operational availability, waste volume reduction, programmatic consistency, and multiple utilization. The results of this study suggest that there are a number of alternative process configurations that are suitable for the treatment of spent ion exchange resin. The determinative evaluation parameters were economic variables (total life-cycle cost or capital cost) and waste volume reduction. Immobilization processes are generally poor in volume reduction. Thermal volume reduction processes tend to have high capital costs. There are immobilization processes and thermal volume reduction processes that can treat all classifications of spent ion exchange resin likely to be encountered. 40 refs., 19 figs., 17 tabs.

  14. Dry halide method for separating the components of spent nuclear fuels

    DOE Patents [OSTI]

    Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

    1998-06-30

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

  15. Spent fuel accumulations and arisings in 'fuel user' states: implications for a world nuclear partnership

    SciTech Connect (OSTI)

    Lineberry, M.J.

    2007-07-01

    Consider the question: In a GNEP world, what are the implications of current spent fuel inventories and possible future accumulations in 'fuel user' states? The inventory today ({approx}95,000 MTHM) and the growth rate ({approx}4,000 MTHM/yr) in likely fuel user states is roughly twice that in the U.S., and thus the problem is substantial. The magnitude of the issue increases if, as is expected, fuel user states grow their nuclear capacity. For purposes of illustration, 2% annual growth assumed in this study. There is an implied imperative to remove spent fuel inventories from fuel user states in some reasonable time (if the spent fuel backlog cannot be cleared by 2100, what motivation is there for GNEP today?). Clearing the backlog requires massive reprocessing capacity, but that makes no economic or strategic sense until the advanced burner reactors are developed and deployed. Thus, burner reactor development is the pacing item with any practical GNEP schedule. With regard to economics, cost estimates for reprocessing (which drive key GNEP costs) are much too disparate today. The disparities must be lessened in order to permit rational decision-making. (author)

  16. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the worlds first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  17. Dry halide method for separating the components of spent nuclear fuels

    DOE Patents [OSTI]

    Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

    1998-01-01

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

  18. Surrogate/spent fuel sabotage aerosol ratio testing:phase 1 summary and results.

    SciTech Connect (OSTI)

    Vigil, Manuel Gilbert; Sorenson, Ken Bryce; Lange, F. , Germany); Nolte, O. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Koch, W. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Dickey, Roy R.; Yoshimura, Richard Hiroyuki; Molecke, Martin Alan; Autrusson, Bruno (Institut de Radioprotection et de Surete Nucleaire , France); Young, F. I.; Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und reaktorsicherheit , Germany)

    2005-10-01

    This multinational test program is quantifying the aerosol particulates produced when a high energy density device (HEDD) impacts surrogate material and actual spent fuel test rodlets. The experimental work, performed in four consecutive test phases, has been in progress for several years. The overall program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This program also provides significant political benefits in international cooperation for nuclear security related evaluations. The spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC), and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission. This report summarizes the preliminary, Phase 1 work performed in 2001 and 2002 at Sandia National Laboratories and the Fraunhofer Institute, Germany, and documents the experimental results obtained, observations, and preliminary interpretations. Phase 1 testing included: performance quantifications of the HEDD devices; characterization of the HEDD or conical shaped charge (CSC) jet properties with multiple tests; refinement of the aerosol particle collection apparatus being used; and, CSC jet-aerosol tests using leaded glass plates and glass pellets, serving as representative brittle materials. Phase 1 testing was quite important for the design and performance of the following Phase 2 test program and test apparatus.

  19. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  20. Calculated and measured drift closure during the spent-fuel test in Climax granite

    SciTech Connect (OSTI)

    Yow, J.L. Jr.; Butkovich, T.R.

    1982-04-01

    Horizontal and vertical measurements of drift closures have been made with a manually operated tape extensometer since about 6 weeks after the emplacement of the spent fuel at various locations along the length of the drifts. The averaged closures are less than 0.6 mm from the onset of measurements through about two years after the spent fuel emplacement. These results have been compared with thermo-elastic finite element calculations using measured medium properties. The comparisons show that most of the closure of the drifts occurred between the time the spent fuel was emplaced and the time of first measurement. The comparisons show that the results track each other, in that where closure followed by dilation is measured, the calculations also show this effect. The agreement is excellent, although where closures of less than 0.2 mm are measured the comparison with calculations is limited by measurement reproducability. Once measurements commenced the averaged measured closures remain to within 30% of the calculated total closure in each drift. 9 figures, 1 table.

  1. Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Lawrence, L.A.

    1995-10-18

    A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford

  2. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    SciTech Connect (OSTI)

    Not Available

    1990-03-01

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs.

  3. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    SciTech Connect (OSTI)

    Hoover, Andrew S; Rudy, Cliff R; Tobin, Steve J; Charlton, William S; Stafford, A; Strohmeyer, D; Saavadra, S

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  4. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    SciTech Connect (OSTI)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  5. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  6. Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

    SciTech Connect (OSTI)

    M.R. Scott

    2005-06-01

    An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, these attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.

  7. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM`S program specifications. (CBS)

  8. DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-06-04

    This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

  9. Effect of Water Radiolysis Caused by Dispersed Radionuclides on Oxidative Dissolution of Spent Fuel in a Final Repository

    SciTech Connect (OSTI)

    Liu Jinsong; Neretnieks, Ivars

    2001-08-15

    When released out of a canister, the radionuclides originally incorporated in the spent fuel can still deposit radiation energy (even more efficiently) into the pore water, cause water radiolysis, and produce oxidants in the buffering material. This phenomenon is termed secondary water radiolysis. The oxidants thus produced can possibly diffuse back to oxidize the spent fuel and to increase the oxidative dissolution rate of the fuel.The effect of the secondary water radiolysis has been identified and preliminarily addressed by a mass-balance model. To explore whether the effect is significant on spent-fuel dissolution, the upper-boundary limit of the effect has been set up by considering a scenario that is very unlikely to occur. Several extreme assumptions have been made: First, the canister fails completely 10{sup 3} yr after deposition; second, the spent fuel is oxidized instantaneously; and third, the radionuclides considered are those that dominantly contribute to radiolysis between 10{sup 3} to 10{sup 5} yr. With these assumptions, the spent-fuel dissolution rate can be increased dramatically if 10% or more of the oxidants produced by the secondary water radiolysis diffuse back to oxidize the spent fuel. It thus indicates that the effect of the secondary water radiolysis could be significant with some extreme assumptions. With more realistic assumptions, the effect could possibly become minimal. The subject is worth further investigation.

  10. COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

    SciTech Connect (OSTI)

    Lombardo, N.J.; Cuta, J.M.; Michener, T.E.; Rector, D.R.; Wheeler, C.L.

    1986-12-01

    This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems.

  11. Spent Fuel Test-Climax: technical measurements data management system description and data presentation

    SciTech Connect (OSTI)

    Carlson, R.C.

    1985-08-01

    The Spent Fuel Test-Climax (SFT-C) was located 420 m below surface in the Climax Stock granite on the Nevada Test Site. The test was conducted under the technical direction of the Lawrence Livermore National Laboratory (LLNL) as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) for the US Department of Energy. Eleven canisters of spent nuclear reactor fuel were emplaced, along with six electrical simulators, in April-May 1980. The spent fuel canisters were retrieved and the electrical simulators de-energized in March-April 1983. During the test, just over 1000 MW-hr of thermal energy was deposited in the site, causing temperature changes 100{sup 0}C near the canisters, and about 5{sup 0} in the tunnels. More than 900 channels of geotechnical, seismological, and test status data were recorded on nearly continuous basis for about 3-1/2 years, ending in September 1983. Most geotechnical instrumentation was known to be temperature sensitive, and thus would require temperature compensation before interpretation. Accordingly, a 10-in. reel of digital tape was off-loaded and shipped to Livermore every 4 to 8 weeks, where the data were verified, organized into 45 one-million-word files, and temperature corrected. The purpose of this report is to document the receipt and processing of the data by LLNL Livermore personnel, present facts about the history of the instruments which may be important to the interpretation of the data, present the data themselves in graphical form for each instrument over its operating lifetime, document the forms and locations in which the data will be archived, and offer the data to the geotechnical community for future use in understanding and predicting the effects of the storage of heat-generating waste in hard rocks such as granite.

  12. Effects of spent fuel types on offsite consequences of hypothetical accidents

    SciTech Connect (OSTI)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-02-18

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor ({chi}/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only {sup 134}Cs, {sup 137}Cs-{sup 137m}Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents.

  13. Truck and rail charges for shipping spent fuel and nuclear waste

    SciTech Connect (OSTI)

    McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

    1986-06-01

    The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials.

  14. Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

    2009-09-01

    During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

  15. Boraflex panel degradation in spent-fuel storage racks at the South Texas Project

    SciTech Connect (OSTI)

    Hoppes, D.F.

    1996-12-31

    Blackness (neutron absorption) testing was conducted in August 1994 on selected South Texas Project (STP) electric generating station spent-fuel pool (SFP) storage racks as required by the surveillance monitoring program. The tests were performed to determine if gaps had developed in the Boraflex neutron poison material and to determine size and location of any gaps identified. The testing was performed by HOLTEC International using a specially designed logging tool containing a {sup 252}Cf neutron source and four boron trifluoride (BF{sub 3}) thermal neutron detectors.

  16. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

  17. Technical Basis Spent Nuclear Fuel (SNF) Project Radiation and Contamination Trending Program

    SciTech Connect (OSTI)

    ELGIN, J.C.

    2000-10-02

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas.

  18. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1996-12-01

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens.

  19. Thermal analysis for a spent reactor fuel storage test in granite

    SciTech Connect (OSTI)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  20. Performance Assessment Analyses Unique to Department of Energy Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Loo, Henry Hung Yiu; Duguid, J. O.

    2000-06-01

    This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented.

  1. Performance assessment analyses unique to Department of Energy spent nuclear fuel

    SciTech Connect (OSTI)

    H. H. Loo; J. O. Duguid

    2000-06-04

    This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented.

  2. DOE Spent Nuclear Fuel Group in Support of Criticality, DBE, TSPA-LA

    SciTech Connect (OSTI)

    Henry Loo

    2000-05-01

    This report presents the basis for grouping the over 250 Department of Energy (DOE) spent nuclear fuel (SNF) types in support of analyses for final repository disposal. For each of the required analyses, the parameters needed in conducting the analyses were identified and reviewed. The grouping proposed for the three types of analyses (criticality, design basis events, and total system performance assessment) are based on the similarities of DOE SNF as a function of these parameters. As necessary, further justifications are provided to further reduce the DOE SNF grouping in support of the Office of Civilian Radioactive Waste Management Systems preclosure and postclosure safety cases.

  3. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

  4. Process and apparatus for generating elemental sulfur and re-usable metal oxide from spent metal sulfide sorbents

    DOE Patents [OSTI]

    Ayala, Raul E. (Clifton Park, NY); Gal, Eli (Lititz, PA)

    1995-01-01

    A process and apparatus for generating elemental sulfur and re-usable metal oxide from spent metal-sulfur compound. Spent metal-sulfur compound is regenerated to re-usable metal oxide by moving a bed of spent metal-sulfur compound progressively through a single regeneration vessel having a first and second regeneration stage and a third cooling and purging stage. The regeneration is carried out and elemental sulfur is generated in the first stage by introducing a first gas of sulfur dioxide which contains oxygen at a concentration less than the stoichiometric amount required for complete oxidation of the spent metal-sulfur compound. A second gas containing sulfur dioxide and excess oxygen at a concentration sufficient for complete oxidation of the partially spent metal-sulfur compound, is introduced into the second regeneration stage. Gaseous sulfur formed in the first regeneration stage is removed prior to introducing the second gas into the second regeneration stage. An oxygen-containing gas is introduced into the third cooling and purging stage. Except for the gaseous sulfur removed from the first stage, the combined gases derived from the regeneration stages which are generally rich in sulfur dioxide and lean in oxygen, are removed from the regenerator as an off-gas and recycled as the first and second gas into the regenerator. Oxygen concentration is controlled by adding air, oxygen-enriched air or pure oxygen to the recycled off-gas.

  5. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  6. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect (OSTI)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  7. Dry-vault storage of spent fuel at the CASCAD facility

    SciTech Connect (OSTI)

    Baillif, L.; Guay, M.

    1989-01-01

    A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooled by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.

  8. Conditioning and Repackaging of Spent Radioactive Cs-137 and Co-60 Sealed Sources in Egypt - 13490

    SciTech Connect (OSTI)

    Hasan, M.A.; Selim, Y.T.; El-Zakla, T.

    2013-07-01

    Radioactive Sealed sources (RSSs) are widely use all over the world in medicine, agriculture, industry, research, etc. The accidental misuse and exposure to RSSs has caused significant environmental contamination, serious injuries and many deaths. The high specific activity of the materials in many RSSs means that the spread of as little as microgram quantities can generate significant risk to human health and inhibit the use of buildings and land. Conditioning of such sources is a must to protect humans and environment from the hazard of ionizing radiation and contamination. Conditioning is also increase the security of these sources by decreasing the probability of stolen and/or use in terrorist attacks. According to the law No.7/2010, Egyptian atomic energy authority represented in the hot laboratories and waste management center (centralized waste facility, HLWMC) has the responsibility of collecting, conditioning, storing and management of all types of radioactive waste from all Egyptian territory including spent radioactive sealed sources (SRSSs). This paper explains the conditioning procedures for two of the most common SRSSs, Cs{sup 137} and Co{sup 60} sources which make up more than 90% of the total spent radioactive sealed sources stored in our centralized waste facility as one of the major activities of hot laboratories and waste management center. Conditioning has to meet three main objectives, be acceptable for storage, enable their safe transport, and comply with disposal requirements. (authors)

  9. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Morrow, Charles W.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level - 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  10. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    SciTech Connect (OSTI)

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Dodge, R.E.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.

  11. Automated Characterization of Spent Fuel through the Multi-Isotope Process (MIP) Monitor

    SciTech Connect (OSTI)

    Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2012-07-31

    This research developed an algorithm for characterizing spent nuclear fuel (SNF) samples based on simulated gamma spectra. The gamma spectra for a variety of light water reactor fuels typical of those found in the United States were simulated. Fuel nuclide concentrations were simulated in ORIGEN-ARP for 1296 fuel samples with a variety of reactor designs, initial enrichments, burn ups, and cooling times. The results of the ORIGEN-ARP simulation were then input to SYNTH to simulate the gamma spectrum for each sample. These spectra were evaluated with partial least squares (PLS)-based multivariate analysis methods to characterize the fuel according to reactor type (pressurized or boiling water reactor), enrichment, burn up, and cooling time. Characterizing some of the features in series by using previously estimated features in the prediction greatly improves the performance. By first classifying the spent fuel reactor type and then using type-specific models, the prediction error for enrichment, burn up, and cooling time improved by a factor of two to four. For some features, the prediction was further improved by including additional information, such as including the predicted burn up in the estimation of cooling time. The optimal prediction flow was determined based on the simulated data. A PLS discriminate analysis model was developed which perfectly classified SNF reactor type. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE.

  12. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    SciTech Connect (OSTI)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

  13. Experience and Lessons Learned from Conditioning of Spent Sealed Sources in Singapore - 13107

    SciTech Connect (OSTI)

    Hong, Dae-Seok; Kang, Il-Sik; Jang, Kyung-Duk; Jang, Won-Hyuk; Hoo, Wee-Teck

    2013-07-01

    In 2010, IAEA requested KAERI (Korea Atomic Energy Research Institute) to support Singapore for conditioning spent sealed sources. Those that had been used for a lightning conductor, check source, or smoke detector, various sealed sources had been collected and stored by the NEA (National Environment Agency) in Singapore. Based on experiences for the conditioning of Ra-226 sources in some Asian countries since 2000, KAERI sent an expert team to Singapore for the safe management of spent sealed sources in 2011. As a result of the conditioning, about 575.21 mCi of Am-241, Ra-226, Co-60, and Sr-90 were safely conditioned in 3 concrete lining drums with the cooperation of the KAERI expert team, the IAEA supervisor, the NEA staff and local laborers in Singapore. Some lessons were learned during the operation: (1) preparations by a local authority are very helpful for an efficient operation, (2) a preliminary inspection by an expert team is helpful for the operation, (3) brief reports before and after daily operation are useful for communication, and (4) a training opportunity is required for the sustainability of the expert team. (authors)

  14. Management of Spent and Disused Sealed Radioactive Sources in the Czech Republic - 12124

    SciTech Connect (OSTI)

    Podlaha, J.

    2012-07-01

    The Czech Republic is a country with a well-developed peaceful utilization of nuclear energy and ionizing radiation. Sealed Radioactive Sources (further also SRS) are broadly used in many areas in the Czech Republic, e.g. in research, industry, medicine, education, agriculture, etc. Legislation in the field of ionizing radiation source utilization has been fully harmonized with European Community legislation. SRS utilization demands a proper system which must ensure the safe use of SRS, including the management of disused (spent) and orphaned SRS. In the Czech Republic, a comprehensive system of SRS management has been established that is comparable with systems in other developed countries. The system covers both legal and institutional aspects. The Central Register of Ionizing Radiation Sources is an important part of the system. It is a tracking system that covers all activities related to SRS, from their production or import to the end of their use (recycling or disposal). Many spent SRS are recycled and can be used for other purposes after inspection, repacking or reprocessing. When the disused SRS are not intended for further use, they are managed as radioactive waste (RAW). The system of SRS management also ensures the suitable resolution of situations connected with improper SRS handling (in the case of orphaned sources, accidents, etc.). (author)

  15. Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Lefebvre, Robert A; Peplow, Douglas E.; Williams, Mark L; Scaglione, John M

    2014-01-01

    The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

  16. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    SciTech Connect (OSTI)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  17. EIS-0251: Department of the Navy Final Environmental Impact Statement for a Container System for the Management of Naval Spent Nuclear Fuel (November 1996)

    Broader source: Energy.gov [DOE]

    This Final Environmental Impact Statement addresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination.

  18. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    SciTech Connect (OSTI)

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

  19. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, J.P.; Miller, W.E.

    1987-11-05

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

  20. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, John P.; Miller, William E.

    1989-01-01

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

  1. HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048

    SciTech Connect (OSTI)

    Halstead, Robert J.; Dilger, Fred

    2003-02-27

    No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

  2. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  3. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  4. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, P.; Dollard, W.J.

    1991-01-08

    An electrochemical apparatus is made having a generator section containing axially elongated electrochemical cells, a fresh gaseous feed fuel inlet, a gaseous feed oxidant inlet, and at least one gaseous spent fuel exit channel, where the spent fuel exit channel passes from the generator chamber to combine with the fresh feed fuel inlet at a mixing apparatus, reformable fuel mixture channel passes through the length of the generator chamber and connects with the mixing apparatus, that channel containing entry ports within the generator chamber, where the axis of the ports is transverse to the fuel electrode surfaces, where a catalytic reforming material is distributed near the reformable fuel mixture entry ports. 2 figures.

  5. MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS A CHALLENGING PROBLEM IN NATURAL CONVECTION

    SciTech Connect (OSTI)

    Fort, James A.; Cuta, Judith M.; Bajwa, C.; Baglietto, E.

    2010-07-18

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10-15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

  6. Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario

    SciTech Connect (OSTI)

    Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

    2007-01-01

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some components heated up beyond their service temperatures, the staff determined that there would be no significant release as a result of the fire for the NAC LWT and similar casks.

  7. Standardized DOE Spent Nuclear Fuel Canister and Transportation System for Shipping to the National Repository

    SciTech Connect (OSTI)

    Pincock, David Lynn; Morton, Dana Keith; Lengyel, Arpad Leslie

    2001-02-01

    The U.S.Department of Energys (DOE) National Spent Nuclear Fuel Program (NSNFP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), has been chartered with the responsibility for developing spent nuclear fuel (SNF) standardized canisters and a transportation cask system for shipping DOE SNF to the national repository. The mandate for this development is outlined in the Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste that states, EM shall design and fabricate DOE SNF canisters for shipment to RW. (1) It also states, EM shall be responsible for the design, NRC certification, and fabrication of the transportation cask system for DOE SNF canisters or bare DOE SNF in accordance with 10 CFR Part 71. (2) In fulfillment of these requirements, the NSNFP has developed four SNF standardized canister configurations and has conceptually designed a versatile transportation cask system for shipping the canisters to the national repository.1 The standardized canister sizes were derived from the national repository waste package design for co-disposal of SNF with high-level waste (HLW). One SNF canister can be placed in the center of the waste package or one can be placed in one of five radial positions, replacing a HLW canister. The internal cavity of the transportation cask was derived using the same logic, matching the size of the internal cavity of the waste package. The size of the internal cavity for the transportation cask allows the shipment of multiple canister configurations with the application of a removable basket design. The standardized canisters have been designed to be loaded with DOE SNF, placed into interim storage, shipped to the national repository, and placed in a waste package without having to be reopened. Significant testing has been completed that clearly demonstrates that the standardized canisters can safely achieve their intended design goals. The transportation cask system will include all of the standard design features, with the addition of dual containment for the shipment of failed fuel. The transportation cask system will also meet the rigorous licensing requirements of the Nuclear Regulatory Commission (NRC) to ensure that the design and the methods of fabrication employed will result in a shipping cask that will safely contain the radioactive materials under all credible accident scenarios. The standardization of the SNF canisters and the versatile design of the transportation cask system will eliminate a proliferation of designs and simplify the operations at the user sites and the national repository.

  8. Comparison and Analysis of Regulatory and Derived Requirements for Certain DOE Spent Nuclear Fuel Shipments; Lessons Learned for Future Spent Fuel Transportation Campaigns

    SciTech Connect (OSTI)

    Kramer, George L., Ph.D.; Fawcett, Rick L.; Rieke, Philip C.

    2003-02-27

    Radioactive materials transportation is stringently regulated by the Department of Transportation and the Nuclear Regulatory Commission to protect the public and the environment. As a Federal agency, however, the U.S. Department of Energy (DOE) must seek State, Tribal and local input on safety issues for certain transportation activities. This interaction has invariably resulted in the imposition of extra-regulatory requirements, greatly increasing transportation costs and delaying schedules while not significantly enhancing the level of safety. This paper discusses the results an analysis of the regulatory and negotiated requirements established for a July 1998 shipment of spent nuclear fuel from foreign countries through the west coast to the Idaho National Engineering and Environmental Laboratory (INEEL). Staff from the INEEL Nuclear Materials Engineering and Disposition Department undertook the analysis in partnership with HMTC, to discover if there were instances where requirements derived from stakeholder interactions duplicate, contradict, or otherwise overlap with regulatory requirements. The study exhaustively lists and classifies applicable Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations. These are then compared with a similarly classified list of requirements from the Environmental Impact Statements (EIS) and those developed during stakeholder negotiations. Comparison and analysis reveals numerous attempts to reduce transportation risk by imposing more stringent safety measures than those required by DOT and NRC. These usually took the form of additional inspection, notification and planning requirements. There are also many instances of overlap with, and duplication of regulations. Participants will gain a greater appreciation for the need to understand the risk-oriented basis of the radioactive materials regulations and their effectiveness in ensuring safety when negotiating extra-regulatory requirements.

  9. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Forsberg, C.W.

    1992-03-24

    A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

  10. Characterization and use of the spent beam for serial operation of LCLS

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Boutet, Sébastien; Foucar, Lutz; Barends, Thomas R. M.; Botha, Sabine; Doak, R. Bruce; Koglin, Jason E.; Messerschmidt, Marc; Nass, Karol; Schlichting, Ilme; Seibert, M. Marvin; et al

    2015-04-11

    X-ray free-electron laser sources such as the Linac Coherent Light Source offer very exciting possibilities for unique research. However, beam time at such facilities is very limited and in high demand. This has led to significant efforts towards beam multiplexing of various forms. One such effort involves re-using the so-called spent beam that passes through the hole in an area detector after a weak interaction with a primary sample. This beam can be refocused into a secondary interaction region and used for a second, independent experiment operating in series. The beam profile of this refocused beam was characterized for amore » particular experimental geometry at the Coherent X-ray Imaging instrument at LCLS. A demonstration of this multiplexing capability was performed with two simultaneous serial femtosecond crystallography experiments, both yielding interpretable data of sufficient quality to produce electron density maps.« less

  11. Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

    SciTech Connect (OSTI)

    Marsha Keister; Kathryn McBride

    2006-08-01

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challengeto develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned.

  12. Precursors for the Immobilization of Radioactive Cesium and Strontium from Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Ortega, Luis H.; McDeavitt, Sean M.

    2007-07-01

    Next generation processes for the recycling of spent nuclear fuel are being developed by the US Department of Energy; this includes solvent extraction methods developed to isolate Cs and Sr fission products and immobilize them in a stable storage form. In this study, simulated (i.e., non-radioactive) cesium and strontium bearing liquid wastes were created and treated with carbon, silica, and alumina at 700 deg. C. The overall goal of this 2006 Nuclear Energy Research Initiative (NERI) project is to synthesize candidate ceramics for Cs and Sr storage and characterize their behavior. The initial results described here confirm the formation of stable compounds resembling strontianite (SrCO{sub 3}) and cesium-aluminosilicate (CsAlSi{sub 2}O{sub 4}). (authors)

  13. Use of Integrated Decay Heat Limits to Facilitate Spent Nuclear Fuel Loading to Yucca Mountain

    SciTech Connect (OSTI)

    Li, Jun; Yim, Man-Sung; McNelis, David; Piet, Steven

    2007-07-01

    As an alternative to the use of the linear loading or areal power density (APD) concept, using integrated decay heat limits based on the use of mountain-scale heat transfer analysis is considered to represent the thermal impact from the deposited spent nuclear fuel (SNF) to the Yucca Mountain repository. Two different integrated decay heat limits were derived to represent both the short-term (up to 50 years from the time of repository closure) and the long-term decay heat effect (up to 1500 years from the time of repository closure). The derived limits were found to appropriately represent the drift wall temperature limit (200 deg. C) and the midway between adjacent drifts temperature limit (96 deg. C) as long as used fuel is uniformly loaded into the mountain. These limits can be a useful practical guide to facilitate the loading of used fuel into Yucca Mountain. (authors)

  14. Numerical approach for the voloxidation process of an advanced spent fuel conditioning process (ACP)

    SciTech Connect (OSTI)

    Park, Byung Heung; Jeong, Sang Mun; Seo, Chung-Seok

    2007-07-01

    A voloxidation process is adopted as the first step of an advanced spent fuel conditioning process in order to prepare the SF oxide to be reduced in the following electrolytic reduction process. A semi-batch type voloxidizer was devised to transform a SF pellet into powder. In this work, a simple reactor model was developed for the purpose of correlating a gas phase flow rate with an operation time as a numerical approach. With an assumption that a solid phase and a gas phase are homogeneous in a reactor, a reaction rate for an oxidation was introduced into a mass balance equation. The developed equation can describe a change of an outlet's oxygen concentration including such a case that a gas flow is not sufficient enough to continue a reaction at its maximum reaction rate. (authors)

  15. Method for processing aluminum spent potliner in a graphite electrode arc furnace

    DOE Patents [OSTI]

    O'Connor, William K.; Turner, Paul C.; Addison, G.W. (AJT Enterprises, Inc.)

    2002-12-24

    A method of processing spent aluminum pot liner containing carbon, cyanide compositions, fluorides and inorganic oxides. The spend aluminum pot liner is crushed, iron oxide is added to form an agglomerated material. The agglomerated material is melted in an electric arc furnace having the electrodes submerged in the molten material to provide a reducing environment during the furnace operation. In the reducing environment, pot liner is oxidized while the iron oxides are reduced to produce iron and a slag substantially free of cyanide compositions and fluorides. An off-gas including carbon oxides and fluorine is treated in an air pollution control system with an afterburner and a scrubber to produce NaF, water and a gas vented to the atmosphere free of cyanide compositions, fluorine, and CO.

  16. Down Select Report of Chemical Hydrogen Storage Materials, Catalysts, and Spent Fuel Regeneration Processes

    SciTech Connect (OSTI)

    Ott, Kevin; Linehan, Sue; Lipiecki, Frank; Aardahl, Christopher L.

    2008-08-24

    The DOE Hydrogen Storage Program is focused on identifying and developing viable hydrogen storage systems for onboard vehicular applications. The program funds exploratory research directed at identifying new materials and concepts for storage of hydrogen having high gravimetric and volumetric capacities that have the potential to meet long term technical targets for onboard storage. Approaches currently being examined are reversible metal hydride storage materials, reversible hydrogen sorption systems, and chemical hydrogen storage systems. The latter approach concerns materials that release hydrogen in endothermic or exothermic chemical bond-breaking processes. To regenerate the spent fuels arising from hydrogen release from such materials, chemical processes must be employed. These chemical regeneration processes are envisioned to occur offboard the vehicle.

  17. Characterization and use of the spent beam for serial operation of LCLS

    SciTech Connect (OSTI)

    Boutet, Sbastien; Foucar, Lutz; Barends, Thomas R. M.; Botha, Sabine; Doak, R. Bruce; Koglin, Jason E.; Messerschmidt, Marc; Nass, Karol; Schlichting, Ilme; Seibert, M. Marvin; Shoeman, Robert L.; Williams, Garth J.

    2015-04-11

    X-ray free-electron laser sources such as the Linac Coherent Light Source offer very exciting possibilities for unique research. However, beam time at such facilities is very limited and in high demand. This has led to significant efforts towards beam multiplexing of various forms. One such effort involves re-using the so-called spent beam that passes through the hole in an area detector after a weak interaction with a primary sample. This beam can be refocused into a secondary interaction region and used for a second, independent experiment operating in series. The beam profile of this refocused beam was characterized for a particular experimental geometry at the Coherent X-ray Imaging instrument at LCLS. A demonstration of this multiplexing capability was performed with two simultaneous serial femtosecond crystallography experiments, both yielding interpretable data of sufficient quality to produce electron density maps.

  18. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    SciTech Connect (OSTI)

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  19. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

    2012-10-22

    CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

  20. AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

    SciTech Connect (OSTI)

    Dewes, J.

    2014-02-24

    The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

  1. Overview of mineral waste form development for the electrometallurgical treatment of spent nuclear fuel

    SciTech Connect (OSTI)

    Pereira, C.; Lewis, M.A.; Ackerman, J.P.

    1996-05-01

    Argonne is developing a method to treat spent nuclear fuel in a molten salt electrorefiner. Wastes from this treatment will be converted into metal and mineral forms for geologic disposal. A glass-bonded zeolite is being developed to serve as the mineral waste form that will contain the fission products that accumulate in the electrorefiner salt. Fission products are ion exchanged from the salt into the zeolite A structure. The crystal structure of the zeolite after ion exchange is filled with salt ions. The salt-loaded zeolite A is mixed with glass frit and hot pressed. During hot pressing, the zeolite A may be converted to sodalite which also retains the waste salt. The glass-bonded zeolite is leach resistant. MCC-1 testing has shown that it has a release rate below 1 g/(m{sup 2}day) for all elements.

  2. Synthetic aggregate compositions derived from spent bed materials from fluidized bed combustion and fly ash

    DOE Patents [OSTI]

    Boyle, Michael J.

    1994-01-01

    Cementitious compositions useful as lightweight aggregates are formed from a blend of spent bed material from fluidized bed combustion and fly ash. The proportions of the blend are chosen so that ensuing reactions eliminate undesirable constituents. The blend is then mixed with water and formed into a shaped article. The shaped article is preferably either a pellet or a "brick" shape that is later crushed. The shaped articles are cured at ambient temperature while saturated with water. It has been found that if used sufficiently, the resulting aggregate will exhibit minimal dimensional change over time. The aggregate can be certified by also forming standardized test shapes, e.g., cylinders while forming the shaped articles and measuring the properties of the test shapes using standardized techniques including X-ray diffraction.

  3. 105-K Basin material design basis feed description for spent nuclear fuel project facilities

    SciTech Connect (OSTI)

    Praga, A.N.

    1998-01-08

    Revisions 0 and 0A of this document provided estimated chemical and radionuclide inventories of spent nuclear fuel and sludge currently stored within the Hanford Site`s 105-K Basins. This Revision (Rev. 1) incorporates the following changes into Revision 0A: (1) updates the tables to reflect: improved cross section data, a decision to use accountability data as the basis for total Pu, a corrected methodology for selection of the heat generation basis fee, and a revised decay date; (2) adds section 3.3.3.1 to expand the description of the approach used to calculate the inventory values and explain why that approach yields conservative results; (3) changes the pre-irradiation braze beryllium value.

  4. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  5. Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Mueller, Don; Goluoglu, Sedat; Hollenbach, Daniel F; Fox, Patricia B

    2007-10-01

    The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

  6. Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility

    SciTech Connect (OSTI)

    Dippre, M. A.

    2003-02-25

    A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational time savings, and significantly improved ALARA exposure.

  7. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect (OSTI)

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 keff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  8. DOE Spent Nuclear Fuel Information In Support of TSPA-VA

    SciTech Connect (OSTI)

    A. Brewer; D. Cresap; D. Fillmore; H. Loo; M. Ebner; R. McCormack

    1998-09-01

    RW has started the viability assessment (VA) effort to determine the feasibility of Yucca Mountain as the first geologic repository for spent nuclear fuel (SNF) and high-level waste. One component of the viability assessment will be a total system performance assessment (TSPA), based on the design concept and the scientific data and analysis available, describing the repository's probable behavior relative to the overall system performance standards. Thus, all the data collected from the Exploratory Studies Facility to-date have been incorporated into the latest TSPA model. In addition, the Repository Integration Program, an integrated probabilistic simulator, used in the TSPA has also been updated by Golder Associates Incorporated at December 1997. To ensure that the Department of Energy-owned (DOE-owned) SNF continues to be acceptable for disposal in the repository, it will be included in the TSPA-VA evaluation. A number of parameters are needed in the TSPA-VA models to predict the performance of the DOE-owned SNF materials placed into the potential repository. This report documents all of the basis and/or derivation for each of these parameters. A number of properties were not readily available at the time the TSPA-VA data was requested. Thus, expert judgement and opinion was utilized to determine a best property value. The performance of the DOE-owned SNF will be published as part of the TSPA-VA report. Each DOE site will be collecting better data as the DOE SNF program moves closer to repository license application. As required by the RW-0333P, the National Spent Nuclear Fuel Program will be assisting each site in qualifying the information used to support the performance assessment evaluations.

  9. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    SciTech Connect (OSTI)

    Manson, S.J.; Gianoulakis, S.E.

    1994-02-01

    The structural properties of spent nuclear fuel shipping containers vary as a function of the cask wall temperature. An analysis is performed to determine the effect of a realistic, though bounding, hot day environment on the thermal behavior of spent fuel shipping casks. These results are compared to those which develop under a steady-state application of the prescribed normal thermal conditions of 10CFR71. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by using the steady-state application of the regulatory boundary conditions. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the regulatory condition. This is due to the conservative assumptions present in the ambient conditions used. The analysis demonstrates that diurnal temperature variations which penetrate the cask wall have maxima substantially less than the corresponding temperatures obtained when applying the steady-state regulatory boundary conditions. Therefore, it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the steady-state interpretation of the 10CFR71 normal conditions.

  10. Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L; Flanagan, Michelle

    2013-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.

  11. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    SciTech Connect (OSTI)

    W. Lee Poe, Jr

    1998-10-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is divided into Part 1 that defines time-dependent releases from each regional site, Part 2 that defines transport conditions through the groundwater, and Part 3 that defines transport through surface water and populations using the surface waters for drinking.

  12. CLEAR LIQUOR SCRUBBING WITH ANHYDRITE PRODUCTION

    SciTech Connect (OSTI)

    R.C. SKARUPA; T.R. CAREY

    1998-08-01

    This project is funded by the US Department of Energy's Federal Energy Technology Center (DOE/FETC) under a cost-sharing PRDA with Radian International. The Electric Power Research Institute (EPRI) is providing co-funding and technical oversight. The project is part of FETC's Advanced Power Systems Program, whose mission is to accelerate the commercialization of affordable, high-efficiency, low emission, coal-fueled electric generating technologies. This project was submitted in response to Area 4 of DOE's Mega-PRDA: Advanced High-Performance SO{sub 2} Control Concepts. The goals of this research area are to develop advanced flue gas desulfurization (FGD) processes that achieve greater than 99% SO{sub 2} removal efficiency, are 25% cheaper than commercial FGD systems, and provide a valuable byproduct that will be recycled rather than disposed. Area 4 also included the development of a byproduct process that could be added to FGD systems to produce high value byproducts for reuse rather than disposal.

  13. Big Island Demonstration Project Black Liquor

    Broader source: Energy.gov [DOE]

    This fact sheet summarizes a U.S. Department of Energy Biomass Program research and development project.

  14. RECOVERY OF URANIUM FROM CARBONATE LEACH LIQUORS

    DOE Patents [OSTI]

    Wilson, H.F.

    1958-07-01

    An improved process is described for the recovery of uranium from vanadifrous ores. In the prior art such ores have been digested with alkali carbonate solutions at a pH of less than 10 and then contacted with a strong base anion exchange resin to separate uranium from vanadium. It has been found that if the exchamge resin feed solution has its pH adjusted to the range 10.8 to 11.8, that vanadium adsorption on the resin is markedly decreased and the separation of uranium from the vanadium is thereby improved.

  15. Definition:Black Liquor | Open Energy Information

    Open Energy Info (EERE)

    into paper pulp removing lignin, hemicelluloses and other extractives from the wood to free the cellulose fibers. View in Wikipedia Retrieved from "http:en.openei.orgw...

  16. Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX

    SciTech Connect (OSTI)

    Van den Akker, B.P.; Ahn, J. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720 (United States)

    2013-07-01

    The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)

  17. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  18. Spent fuel storage and waste management fuel cycle optimization using CAFCA

    SciTech Connect (OSTI)

    Brinton, S.; Kazimi, M.

    2013-07-01

    Spent fuel storage modeling is at the intersection of nuclear fuel cycle system dynamics and waste management policy. A model that captures the economic parameters affecting used nuclear fuel storage location options, which complements fuel cycle economic assessment has been created using CAFCA (Code for Advanced Fuel Cycles Assessment) of MIT. Research has also expanded to the study on dependency of used nuclear fuel storage economics, environmental impact, and proliferation risk. Three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module which provides an easy to use interface for education on fuel cycle waste management economic impacts. Storage options costs can be compared to literature values with simple variation available for sensitivity study. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (authors)

  19. MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

    SciTech Connect (OSTI)

    Carbajo, Juan J

    2012-01-01

    Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, cooling to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total decay heat of the fuel in the pool was, at the time of the accident, 2.284 MWt, of which 1.872 MWt were from the 548 assemblies of the last core discharged and 0.412 MWt were from the older 783 assemblies. These decay heat values were calculated at Oak Ridge National Laboratory using the ORIGEN2.2 code (Ref. 2) - they agree with values reported elsewhere (Ref. 3). The pool dimensions are 9.9 m x 12.2 m x 11.8 m (height), and with the water level at 11.5 m, the pool volume is 1389 m3, of which only 1240 m3 is water, as some volume is taken by the fuel and by the fuel racks. The initial water temperature of the SFP4 was assumed to be 301 K. The fuel racks are made of an aluminum alloy but are modeled in MELCOR with stainless steel and B4C. MELCOR calculations were completed for different initial water levels: 11.5 m (pool almost full, water is only 0.3 m below the top rim), 4.4577 m (top of the racks), 4.2 m, and 4.026 m (top of the active fuel). A calculation was also completed for a rapid loss of water due to a leak at the bottom of the pool, with the fuel rapidly uncovered and oxidized in air. Results of these calculations are shown in the enclosed Table I. The calculation with the initial water level at 11.5 m (full pool) takes 11 days for the water to boil down to the top of the fuel racks, 11.5 days for the fuel to be uncovered, 14.65 days to generate 150 kg of hydrogen and 19 days for the pool to be completely dry. The calculation with the initial water level at 4.4577 m, takes 1.1 days to uncover the fuel and 4.17 days to generate 150 kg of hydrogen. The calculation with the initial water level at 4.02 m takes 3.63 days to generate 150 kg of hydrogen this is exactly the time when the actual explosion occurred in Unit 4. Finally, fuel oxidation in air after the pool drained the water in 20 minutes, generates only 10 kg of hydrogen this is because very little steam is available and Zircaloy (Zr) oxidation with the oxygen of the air does not generate hydrogen. MELCOR calculated water levels and hydrogen generated in the SFP4 as a function of time for initial water levels of 4.457 m, 4.2 m and 4.02 m are shown in Figs. 1 and 2. Water levels increase at the beginning due to the expansion of the water during the heat-up from 301 K to 373 K. Boiling occurs after the water temperature reaches 373 K. The total amount of hydrogen generated is ~2000 kg, this amount includes hydrogen generated from Zr, which is the largest amount (~1580 kg), from stainless steel (~360 kg), and from B4C (~60 kg). In theory, it is possible to generate up to 3.4 kg of hydrogen per assembly (from oxidation of Zr in the fuel cladding and box), or a total of 4,525 kg from the hot 1331 assemblies stored in the SFP4. The hydrogen generated from oxidation of steel and B4C will be additional. So the answers to the questions are YES according to these MELCOR calculations, enough hydrogen (150 kg) could be generated in the SFP4 3.64 days after the earthquake to produce ...

  20. Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

    SciTech Connect (OSTI)

    Goodsell, Alison V.; Henzl, Vladimir; Swinhoe, Martyn T.

    2014-04-01

    New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238U first fission contribution was found to not affect the DDA performance in reconstructing 239Pueff. As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

  1. FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405

    SciTech Connect (OSTI)

    Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert

    2013-07-01

    A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

  2. An Analysis of Dual Zone Loading for Shipping Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Allen, William Christopher; Yim, Man-Sung

    2007-07-01

    The bumps current fuel assembly designs can achieve exceeds the fuel assembly burnups the current fleet of shipping casks can ship. One method of handling this situation which has been proposed is regionalized loading. This concept involves administratively separating the fuel basket of a shipping cask into two or more regions and loading fuel with different burnup, cooling times and enrichments into these regions. To evaluate how regionalized loading patterns might affect shipping spent nuclear fuel in comparison to uniform loading, a test case study was performed using fuel assemblies discharged from an actual nuclear plant and a shipping cask licensed by the NRC. Using the same fuel assemblies and shipping cask, results were obtained assuming a uniform loading pattern and compared to the results obtained assuming a dual zone loading pattern. Source terms for the analysis were generated using SAS2 and the dose levels were calculated using MCNPS. The analysis showed that the dual zone loading reduced the amount of time required to ship the given quantity of fuel by roughly thirty percent compared to the uniform loading. The average dose rate to the transportation workers and the public due to the implementation of dual zone loading increased. Implications of these increases are discussed. (authors)

  3. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  4. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    SciTech Connect (OSTI)

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-11-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented.

  5. Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

    SciTech Connect (OSTI)

    Lam, Poh -Sang; Sindelar, Robert L.

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  6. Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

    SciTech Connect (OSTI)

    Lam, P.; Sindelar, R.

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  7. COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

    SciTech Connect (OSTI)

    Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

    2003-02-27

    Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal.

  8. Long-term, low-temperature oxidation of PWR spent fuel: Interim transition report

    SciTech Connect (OSTI)

    Einziger, R.E.; Buchanan, H.C.

    1988-05-01

    Since some of the fuel rods will be breached and eventually most of the cladding will corrode, exposing fuel, one factor influencing the ability of spent fuel to retain radionuclides is its oxidation state in the expected moist air atmosphere. Oxidation of the fuel could split the cladding, exposing additional fuel and changing the leaching characteristics. Thermodynamically, there is no reason why UO{sub 2} should not oxidize completely to UO{sub 3} at repository temperatures. The underlying uncertainty is the rate of oxidation. Extrapolation of higher temperature data indicates that insufficient oxidation to convert all of the fuel to U{sub 3}O{sub 8} will occur during the first 10,000 years. However, lower oxidation states, such as U{sub 4}O{sub 9} and U{sub 3}O{sub 7}, might form. To date, the tests have run between 3200 and 4100 hours out of a planned 16,000-hour duration. Some preliminary conclusions can be drawn: (1) Moisture content of the air has no significant effect on oxidation rate, (2) the data have an uncertainty of 15 to 20%, which must be accounted for in the interpretation of single sample tests, and (3) below 175{degree}C, the oxidation rate is dependent on the particle size in the sample. The smaller particles oxidize more rapidly. 19 refs., 23 figs., 7 tabs.

  9. Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1997-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  10. Spent nuclear fuel project, Cold Vacuum Drying Facility human factors engineering (HFE) analysis: Results and findings

    SciTech Connect (OSTI)

    Garvin, L.J.

    1998-07-17

    This report presents the background, methodology, and findings of a human factors engineering (HFE) analysis performed in May, 1998, of the Spent Nuclear Fuels (SNF) Project Cold Vacuum Drying Facility (CVDF), to support its Preliminary Safety Analysis Report (PSAR), in responding to the requirements of Department of Energy (DOE) Order 5480.23 (DOE 1992a) and drafted to DOE-STD-3009-94 format. This HFE analysis focused on general environment, physical and computer workstations, and handling devices involved in or directly supporting the technical operations of the facility. This report makes no attempt to interpret or evaluate the safety significance of the HFE analysis findings. The HFE findings presented in this report, along with the results of the CVDF PSAR Chapter 3, Hazards and Accident Analyses, provide the technical basis for preparing the CVDF PSAR Chapter 13, Human Factors Engineering, including interpretation and disposition of findings. The findings presented in this report allow the PSAR Chapter 13 to fully respond to HFE requirements established in DOE Order 5480.23. DOE 5480.23, Nuclear Safety Analysis Reports, Section 8b(3)(n) and Attachment 1, Section-M, require that HFE be analyzed in the PSAR for the adequacy of the current design and planned construction for internal and external communications, operational aids, instrumentation and controls, environmental factors such as heat, light, and noise and that an assessment of human performance under abnormal and emergency conditions be performed (DOE 1992a).

  11. Depleted uranium oxides and silicates as spent nuclear fuel waste package fill materials

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

    A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated that uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be filled with SNF and then filled with depleted uranium (DU) ({approximately}0.2 wt % {sup 235}U) dioxide (UO{sub 2}) or DU silicate-glass beads. Fission products and actinides can not escape the SNF UO{sub 2} crystals until the UO{sub 2} dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of AT groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion seals the WP from water flow. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

  12. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study 15134

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  13. DOE Spent Nuclear Fuel Information in Support of TSPA-SR

    SciTech Connect (OSTI)

    H. H. Loo

    1999-08-01

    The Department of Energy (DOE) Office of Civilian Radioactive Waste Management (RW) has started the recommendation (SR) effort to show that Yucca Mountain could be selected as the first geologic repository for spent nuclear fuel (SNF) and high-level waste. One component of the site recommendation will be a total system performance assessment (TSPA), based on the design concept and the scientific data and analysis available, describing the repository's probable behavior relative to the overall system performance standards. Thus, all the data collected from the Exploratory Studies Facilities to-date have been incorporated into the latest TSPA model. To ensure that the DOE-owned SNF continues to be acceptable for disposal in the repository, it will be included in the TSPA-SR evaluation. A number of parameters are needed in the TSPA-SR models to predict the performance of the DOE-owned SNF materials placed into the potential repository. This report documents all of the basis and/or derivation for each of these parameters. A number of properties were not readily available at the time the TSPA-SR data were requested. Thus, expert judgement and opinion were used to determine a best property value. The performance of the DOE-owned SNF will be published as part of the TSPA-SR report.

  14. The MacArthur Maze Fire and Roadway Collapse: A "Worst Case Scenario" for Spent Nuclear Fuel Transportation?

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Easton, Earl P.; Adkins, Harold E.; Cuta, Judith M.; Klymyshyn, Nicholas A.; Suffield, Sarah R.

    2012-07-06

    In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the "MacArthur Maze." The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC's analysis of the consequences of this type of severe accident.

  15. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  16. Selective Separation of Cs and Sr from LiCl-Based Salt for Electrochemical Processing of Oxide Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P Sachdev

    2008-07-01

    Electrochemical processing technology is currently being used for the treatment of metallic spent fuel from the Experimental Breeder Reactor-II at Idaho National Laboratory. The treatment of oxide-based spent nuclear fuel via electrochemical processing is possible provided there is a front-end oxide reduction step. During this reduction process, certain fission products, including Cs and Sr, partition into the salt phase and form chlorides. Both solid state and molten LiCl-zeolite-A ion exchange tests were conducted for selectively removing Cs and Sr from LiCl-based salt. The solid-state tests produced in excess of 99% removal of Cs and Sr. The molten state tests failed due to phase transformation of the zeolite structure when in contact with the molten LiCl salt.

  17. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, Philip (Plum Borough, PA); Dollard, Walter J. (Churchill Borough, PA)

    1991-01-01

    An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

  18. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    SciTech Connect (OSTI)

    Sanders, T.L. ); Jordan, H. . Rocky Flats Plant); Pasupathi, V. ); Mings, W.J. ); Reardon, P.C. )

    1991-09-01

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  19. FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

    SciTech Connect (OSTI)

    Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

    2014-03-31

    This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

  20. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  1. Radioactive Waste Management: Study of Spent Fuel Dissolution Rates in Geological Storage Using Dosimetry Modeling and Experimental Verification

    SciTech Connect (OSTI)

    Hansen, Brady; Miller, William

    2011-10-28

    This research will provide improved predictions into the mechanisms and effects of radiolysis on spent nuclear fuel dissolution in a geological respository through accurate dosimetry modeling of the dose to water, mechanistic chemistry modeling of the resulting radiolytic reactions and confirmatory experimental measurements. This work will combine effort by the Nuclear Science and Engineering Institute (NSEI) and the Missouri University Research Reactor (MURR) at the University of Missouri-Columbia, and the expertise and facilities at the Pacific Northwest National Laboratory (PNNL).

  2. American National Standard: design requirements for light water reactor spent fuel storage facilities at nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1983-10-07

    This standard presents necessary design requirements for facilities at nuclear power plants for the storage and preparation for shipment of spent fuel from light-water moderated and cooled nuclear power stations. It contains requirements for the design of fuel storage pool; fuel storage racks; pool makeup, instrumentation and cleanup systems; pool structure and integrity; radiation shielding; residual heat removal; ventilation, filtration and radiation monitoring systems; shipping cask handling and decontamination; building structure and integrity; and fire protection and communication.

  3. Technical Approach and Plan for Transitioning Spent Nuclear Fuel (SNF) Project Facilities to the Environmental Restoration Program

    SciTech Connect (OSTI)

    SKELLY, W.A.

    1999-10-06

    This document describes the approach and process in which the 100-K Area Facilities are to be deactivated and transitioned over to the Environmental Restoration Program after spent nuclear fuel has been removed from the K Basins. It describes the Transition Project's scope and objectives, work breakdown structure, activity planning, estimated cost, and schedule. This report will be utilized as a planning document for project management and control and to communicate details of project content and integration.

  4. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation

    SciTech Connect (OSTI)

    G. G. Hall

    2000-02-01

    This report presents the results of the 1999 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  5. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation

    SciTech Connect (OSTI)

    Hall, Gregory Graham

    2002-02-01

    This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  6. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation (2005)

    SciTech Connect (OSTI)

    Hall, Gregory Graham

    2001-02-01

    This report presents the results of the 2000 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  7. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island - Unit 2 Independent Spent Fuel Storage Installation

    SciTech Connect (OSTI)

    Gregory G. Hall

    2003-02-01

    This report presents the results of the 2002 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  8. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation

    SciTech Connect (OSTI)

    Hall, Gregory Graham

    2001-02-01

    This report presents the results of the 2000 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  9. Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation

    SciTech Connect (OSTI)

    Fulton, J.C.

    1994-10-01

    Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II.

  10. Draft Supplement Analysis: Two Proposed Shipments of Commercial Spent Nuclear Fuel to Idaho National Laboratory for Research and Development Purposes

    Broader source: Energy.gov [DOE]

    DOE is proposing to transport, in two separate truck shipments, small quantities of commercial power spent nuclear fuel (SNF) to the Idaho National Laboratory (INL) Site for research purposes consistent with the mission of the DOE Office of Nuclear Energy. DOE is preparing a Supplement Analysis to determine whether an existing environmental impact statement should be supplemented, a new environmental impact statement should be prepared, or that no further NEPA documentation is required for this proposed action.

  11. Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel

    Office of Environmental Management (EM)

    National Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders Forum Forum 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting May 11, 2011 May 11, 2011 Evaluation of Shortline Railroads Evaluation of Shortline Railroads & & & & SNF/HLW Rail Shipment Inspections SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel Tasked for the

  12. Simultaneous separation of cesium and strontium from spent nuclear fuel using the fission-product extraction process

    SciTech Connect (OSTI)

    Law, J.D.; Peterman, D.R.; Riddle, C.L.; Meikrantz, D.A.; Todd, T.A.

    2008-07-01

    The Fission-Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Global Nuclear Energy Partnership (GNEP) for the simultaneous separation of cesium and strontium from spent LWR fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository and, when combined with the separation of Am and Cm, could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly-specific extractants: 4,4',(5')-di-(t-butyl-dicyclohexano)- 18-crown-6 (DtBuCH18C6) and calix[4]arene-bis-(t-octyl-benzo-crown-6 ) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium, and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with simulated and actual spent-nuclear-fuel feed solution in centrifugal contactors are detailed. Removal efficiencies, co-extraction of metals, and process hydrodynamic performance ar e discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel. Recent advances in the evaluation of alternative calixarenes with increased solubility and stability are also detailed. (authors)

  13. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect (OSTI)

    S. D. Herrmann; S. X. Li

    2010-09-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  14. EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposed action to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain  for the disposal of spent nuclear fuel and high-level...

  15. Fifth National Report for the Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management

    Broader source: Energy.gov [DOE]

    This Fifth United States of America (U.S.) National Report updates the Fourth Report published in October 2011, under the terms of the Joint Convention on the Safety of Spent Fuel Management and on...

  16. Processes for washing a spent ion exchange bed and for treating biomass-derived pyrolysis oil, and apparatuses for treating biomass-derived pyrolysis oil

    DOE Patents [OSTI]

    Baird, Lance Awender; Brandvold, Timothy A.

    2015-11-24

    Processes and apparatuses for washing a spent ion exchange bed and for treating biomass-derived pyrolysis oil are provided herein. An exemplary process for washing a spent ion exchange bed employed in purification of biomass-derived pyrolysis oil includes the step of providing a ion-depleted pyrolysis oil stream having an original oxygen content. The ion-depleted pyrolysis oil stream is partially hydrotreated to reduce the oxygen content thereof, thereby producing a partially hydrotreated pyrolysis oil stream having a residual oxygen content that is less than the original oxygen content. At least a portion of the partially hydrotreated pyrolysis oil stream is passed through the spent ion exchange bed. Water is passed through the spent ion exchange bed after passing at least the portion of the partially hydrotreated pyrolysis oil stream therethrough.

  17. User`s guide to REVERT. A CDC 7600 program for converting Spent Fuel Test - Climax data to engineering units, with corrections

    SciTech Connect (OSTI)

    Hage, G.

    1984-10-01

    A CDC 7600 computer program, REVERT, can revise Spent Fuel Test - Climax data files using one of several algorithms, depending on the type of data. The algorithms use coefficients from a separate file organized by data type identifiers. REVERT can also make that file of coefficients, using data from tapes made by Hewlett-Packard equipment employed for data acquisition on the spent Fuel Test - Climax at NTS. 12 references.

  18. DOE/EIS-0251; Supplemental Analysis For a Container System for the Management of DOE Spent Nuclear Fuel Located at the INEEL

    Office of Environmental Management (EM)

    ID-10636 SUPPLEMENT ANALYSIS FOR A CONTAINER SYSTEM FOR THE MANAGEMENT OF DOE SPENT NUCLEAR FUEL LOCATED AT THE INEEL March 1999 U.S. Department of Energy Idaho Operations Office Idaho Falls, Idaho DOE/ID-10636 SUPPLEMENT ANALYSIS FOR A CONTAINER SYSTEM FOR THE MANAGEMENT OF DOE SPENT NUCLEAR FUEL LOCATED AT THE INEEL March 1999 Department of Energy Idaho Operations Office Idaho Falls, Idaho SNF Supplement Analysis ii March 1999 CONTENTS Acronyms and Abbreviations

  19. DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

    SciTech Connect (OSTI)

    Sindelar, R.; Deible, R.

    2011-04-27

    The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

  20. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    SciTech Connect (OSTI)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary analysis indicates the mass of 239Pu may be determined with a precision on the order of 5%, without the need for operator-supplied fuel information or operational histories. This paper will present our work to date on this topic, indicate our preliminary findings for a conceptual assay approach, discuss resilience against spoofing, and outline our future plans for evaluating the NRTA technique for SFA plutonium determination.

  1. Expected environments in high-level nuclear waste and spent fuel repositories in salt

    SciTech Connect (OSTI)

    Claiborne, H.C.; Rickertsen, L.D., Graham, R.F.

    1980-08-01

    The purpose of this report is to describe the expected environments associated with high-level waste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.

  2. NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547

    SciTech Connect (OSTI)

    Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H.; Pan, Y.

    2012-07-01

    Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

  3. Passive Tomography for Spent Fuel Verification: Analysis Framework and Instrument Design Study

    SciTech Connect (OSTI)

    White, Timothy A.; Svard, Staffan J.; Smith, Leon E.; Mozin, Vladimir V.; Jansson, Peter; Davour, Anna; Grape, Sophie; Trellue, H.; Deshmukh, Nikhil S.; Wittman, Richard S.; Honkamaa, Tapani; Vaccaro, Stefano; Ely, James

    2015-05-18

    The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information. The second objective is to provide quantitative measures of pin-by-pin properties, e.g. activity of key isotopes or pin attributes such as cooling time and relative burnup, for the detection of anomalies and/or verification of operator-declared data. The efficacy of GET to meet these two verification objectives will be evaluated across a range of fuel types, burnups, and cooling times, and with a target interrogation time of less than 60 minutes. The evaluation of GET viability for safeguards applications is founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types are used to produce simulated tomographer responses to large populations of virtual fuel assemblies. Instrument response data are processed by a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives are defined and used to evaluate the performance of the methods. This paper will provide a description of the analysis framework and evaluation metrics, example performance-prediction results, and describe the design of a universal GET instrument intended to support the full range of verification scenarios envisioned by the IAEA.

  4. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect (OSTI)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  5. Plan for characterization of K Basin Spent Nuclear Fuel and sludge. Revision 1

    SciTech Connect (OSTI)

    Lawrence, L.A.

    1995-10-05

    This plan outlines a Characterization Program that provides the necessary data to support the Integrated Process Strategy scope and schedules for the Spent Nuclear Fuel (SNF) and sludge stored in the Hanford K Basins. The plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years (i.e., FY 1995 through FY 1998) and is limited to in-situ and laboratory examinations of the SNF and sludge in the K East and K West Basins. In order to assure the scope and schedule of the Characterization Program fully supports the Integrated Process Strategy, key project management has approved the plan. The intent of the program is to provide bounding behavior for the fuel, and acceptability for the transfer of the sludge to the Double Shell Tanks. Fuel examinations are based on two shipping compains from the K West Basin and one from the K East Basin with coincident sludge sampling campaings for the associated canister sludge. Sampling of the basin floor and pit sludge will be conducted independent of the fuel and canister sludge shipping activities. Fuel behavior and properties investigated in the laboratory include physical condition, hydride and oxide content, conditioning testing, oxidation kinetics, and dry storage behavior. These laboratory examinations are expected to provide the necessary data to establish or confirm fuel conditioning process limits and support safety analysis. Sludge laboratory examinations include measurement of quantity and content, measurement of properties for equipment design and recovery process limits and support safety analysis. Sludge laboratory examinations include measurement of quantity and content, measurement of properties for equipment design and recovery precesses, tank farm acceptance, simulant development, measurement of corrosion products, and measurements of drying behavior.

  6. Characterization Program Management Plan for Hanford K Basin Spent Nuclear Fuel (SNF) (OCRWM)

    SciTech Connect (OSTI)

    BAKER, R.B.; TRIMBLE, D.J.

    2000-12-12

    The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. The transport and storage of K Basin sludge are affected by recent path forward revisions. These revisions require additional laboratory analyses of the sludge to complete the acquisition of required supporting engineering data. Hence, this revision of the management plan provides the overall work control for these remaining SNF and sludge characterization activities given the current organizational structure of the SNF Project.

  7. Simulated dry storage test of a spent PWR nuclear fuel assembly in air

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.; Oden, D.R.; Stidham, D.L.; Garnier, J.E.; Weeks, D.L.; Dobbins, J.C.

    1985-02-01

    The purpose of the dry storage test was to investigate the behavior of Zircaloy-clad spent fuel in air between 200 and 275/sup 0/C. Atmospheric air was used for the cover gas because of the interest in establishing regimes where air inleakage into an initially inert system would not cause potential fuel degradation. Samples of the cover gas atmosphere were extracted monthly to determine fission gas concentrations as a function of time. The oxygen concentration was monitored to detect oxygen depletion, which would signal oxidation of the fuel. The gas analyses indicated very low but detectable levels of /sup 85/Kr during the first month of the test. A large increase (five orders of magnitude) in /sup 85/Kr and the appearance of helium in the cover gas indicated that a fuel rod had breached during the second month of the test. Stress rupture calculations showed that the stresses and temperatures were too low to expect breaches to form in defect-free cladding. It is theorized that the breach occurred in a fuel rod weakened by an existing cladding or end cap defect. Calculations based on the rate of /sup 85/Kr release suggest that the diameter of the initial breach was about 25 microns. A post-test fuel examination will be performed to locate and investigate the cause of the cladding breach and to determine if detectable fuel degradation progressed after the breach occurred. The post-test evaluation will define the consequences of a fuel rod breach occurring in an air cover gas at 270/sup 0/C, followed by subsequent exposure to air at a prototypic descending temperature.

  8. Evaluation of the Strategic Value of Fully Burnt PBMR Spent Fuel - A Report to ISPO in Response to IAEA Letter Request (2004-08-30)

    SciTech Connect (OSTI)

    A. M. Ougouag; H. D. Gougar; T. A. Todd

    2006-05-01

    The IAEA needs to determine the value of imposing safeguards on the spent fuel storage at the Pebble Bed Modular Reactor (PBMR) planned for construction in the Republic of South Africa. The PBMR will use hundreds of thousands of fuel elements in the shape of small spheres (6 cm in diameter). The PBMR plant design calls for the storage on site of all the spent fuel generated during the whole life of the reactor, expected to span 40 years. The spent fuel storage system is designed (or to be designed) for a functional life of 80 years. If it is determined that the spent fuel contains materials of interest to a would-be proliferant, then safeguards would have to be imposed and maintained until the spent fuel elements are processed into a form and composition that no longer requires safeguards. The problem addressed in this report is the determination of the strategic value of the spent fuel to such a would-be proliferant.

  9. Final environmental impact statement for the construction and operation of an independent spent fuel storage installation to store the Three Mile Island Unit 2 spent fuel at the Idaho National Engineering and Environmental Laboratory. Docket Number 72-20

    SciTech Connect (OSTI)

    1998-03-01

    This Final Environmental Impact Statement (FEIS) contains an assessment of the potential environmental impacts of the construction and operation of an Independent Spent Fuel Storage Installation (ISFSI) for the Three Mile Island Unit 2 (TMI-2) fuel debris at the Idaho National Engineering and Environmental laboratory (INEEL). US Department of Energy-Idaho Operations Office (DOE-ID) is proposing to design, construct, and operate at the Idaho Chemical Processing Plant (ICPP). The TMI-2 fuel debris would be removed from wet storage, transported to the ISFSI, and placed in storage modules on a concrete basemat. As part of its overall spent nuclear fuel (SNF) management program, the US DOE has prepared a final programmatic environmental impact statement (EIS) that provides an overview of the spent fuel management proposed for INEEL, including the construction and operation of the TMI-2 ISFSI. In addition, DOE-ID has prepared an environmental assessment (EA) to describe the environmental impacts associated with the stabilization of the storage pool and the construction/operation of the ISFSI at the ICPP. As provided in NRC`s NEPA procedures, a FEIS of another Federal agency may be adopted in whole or in part in accordance with the procedures outlined in 40 CFR 1506.3 of the regulations of the Council on Environmental Quality (CEQ). Under 40 CFR 1506.3(b), if the actions covered by the original EIS and the proposed action are substantially the same, the agency adopting another agency`s statement is not required to recirculate it except as a final statement. The NRC has determined that its proposed action is substantially the same as actions considered in DOE`s environmental documents referenced above and, therefore, has elected to adopt the DOE documents as the NRC FEIS.

  10. Spent Nuclear Fuel Trasportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

    SciTech Connect (OSTI)

    M. Keister; K, McBride

    2006-08-28

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository (if licensed) in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challenge--to develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned. The objective of this lessons learned study was to identify successful, best-in-class trends and commonalities from past shipping campaigns, which OCRWM could consider when planning for the development and operation of a repository transportation system. Note: this paper is for analytical and discussion purposes only, and is not an endorsement of, or commitment by, OCRWM to follow any of the comments or trends. If OCRWM elects to make such commitments at a future time, they will be appropriately documented in formal programmatic policy statements, plans and procedures. Reviewers examined an extensive study completed in 2003 by DOE's National Transportation Program (NTP), Office of Environmental Management (EM), as well as plans and documents related to SNF shipments since issuance of the NTP report. OCRWM examined specific planning, business, institutional and operating practices that have been identified by DOE, its transportation contractors, and stakeholders as important issues that arise repeatedly. In addition, the review identifies lessons learned or activities/actions which were found not to be productive to the planning and conduct of SNF shipments (i.e., negative impacts). This paper is a 'looking back' summary of lessons learned across multiple transportation campaigns. Not all lessons learned are captured here, and participants in some of the campaigns have divergent opinions and perspectives about which lessons are most critical. This analysis is part of a larger OCRWM benchmarking effort to identify best practices to consider in future transportation of radioactive materials ('looking forward'). Initial findings from this comprehensive benchmarking analysis are expected to be available in late fall 2006.

  11. Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

    SciTech Connect (OSTI)

    Sobes, Vladimir; Scaglione, John M.; Wagner, John C.; Dunn, Michael E.

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

  12. Discussion of Available Methods to Support Reviews of Spent Fuel Storage Installation Cask Drop Evaluations

    SciTech Connect (OSTI)

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections. The Section II provides a summary of the literature search on concrete finite element models. The Section III discusses commercial codes. The Section IV provides recommendations.

  13. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10? cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 Nm was obtained at maximum curvature 4.0 m?. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 Nm. Failure was observed in nine of the tested rod specimens. The cycles to failure were from 1.22 10? to 4.70 10?, when the amplitude varied from 15.2 to 7.6 Nm. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 Nm. The on-line monitoring revealed that the flexural rigidity was a little lower due to the high level of loading, from 25 to 42 Nm. Generally, no substantial change of rigidity was observed based on on-line monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of lifetime with the increasing amplitude is well defined.

  14. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect (OSTI)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

  15. Plan for Characterization of K Basin Spent Nuclear Fuel (SNF) and Sludge (OCRWM)

    SciTech Connect (OSTI)

    TRIMBLE, D.J.

    2000-08-24

    This is an update of the plan for the characterization of spent nuclear fuel (SNF) and sludge stored in the Hanford K West and K East Basins. The purpose of the characterization program is to provide fuel and sludge data in support of the SNF Project in the effort to remove the fuel from the K Basins and place it into dry storage. Characterization of the K Basin fuel and sludge was initiated in 1994 and has been guided by the characterization plans (Abrefah 1994, Lawrence 1995a, Lawrence 1995b) and the characterization program management plan (PMP) (Lawrence 1995c, Lawrence 1998, Trimble 1999). The fuel characterization was completed in 1999. Summaries of these activities were documented by Lawrence (1999) and Suyama (1999). Lawrence (1999) is a summary report providing a road map to the detailed documentation of the fuel characterization. Suyama (1999) provides a basis for the limited characterization sample size as it relates to supporting design limits and the operational safety envelope for the SNF Project. The continuing sludge characterization is guided by a data quality objective (DQO) (Makenas 2000) and a sampling and analysis plan (SAP) (Baker, Welsh and Makenas 2000) The original intent of the characterization program was ''to provide bounding behavior for the fuel'' (Lawrence 1995a). To accomplish this objective, a fuel characterization program was planned that would provide data to augment data from the literature. The program included in-situ examinations of the stored fuel and laboratory testing of individual elements and small samples of fuel (Lawrence 1995a). Some of the planned tests were scaled down or canceled due to the changing needs of the SNF Project. The fundamental technical basis for the process that will be used to place the K Basin fuel into dry storage was established by several key calculations. These calculations characterized nominal and bounding behavior of fuel in Multi-Canister Overpacks (MCOs) during processing and storage. Characterization data supported input parameters that were used for the calculations. These parameters are documented in the Project Technical Databook (Reilly 1998). Calculations and reviews used in developing the parameters describe how characterization data were used in the parameter development, e.g. statistically based, added factors of conservatism, etc.

  16. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and at cask-loading-specific conditions could be performed to demonstrate that release is within the allowable leak rates of the cask.

  17. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    SciTech Connect (OSTI)

    IRWIN, J.J.

    2000-11-18

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed from the MCO back to the K Basins.

  18. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    SciTech Connect (OSTI)

    Manson, S.J.; Gianoulakis, S.E.

    1994-04-01

    An examination of the effect of a realistic (though conservative) hot day environment on the thermal transient behavior of spent fuel shipping casks is made. These results are compared to those that develop under the prescribed normal thermal condition of 10 CFR 71. Of specific concern are the characteristics of propagating thermal waves, which are set up by diurnal variations of temperature and insolation in the outdoor environment. In order to arrive at a realistic approximation of these variations on a conservative hot day, actual temperature and insolation measurements have been obtained from the National Climatic Data Center (NCDC) for representatively hot and high heat flux days. Thus, the use of authentic meteorological data ensures the realistic approach sought. Further supporting the desired realism of the modeling effort is the use of realistic cask configurations in which multiple laminations of structural, shielding, and other materials are expected to attenuate the propagating thermal waves. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by enforcement of the regulatory environmental conditions of 10 CFR 71. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the prescribed regulatory conditions. However, the temperature differences are small enough that the normal conservative assumptions that are made in the course of typical cask evaluations should correct for any potential violations. The analysis demonstrates that diurnal temperature variations that penetrate the cask wall all have maxima substantially less than the corresponding regulatory solutions. Therefore it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the conditions of 10 CFR 71.

  19. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 1 Fuel

    SciTech Connect (OSTI)

    PACKER, M.J.

    1999-11-04

    Metallic uranium Spent Nuclear Fuel (SNF) is currently stored within two water filled pools, 105-KE Basin (KE Basin) and 105-KW Basin (KW Basin), at the United States Department of Energy (U.S. DOE) Hanford Site, in southeastern Washington State. The Spent Nuclear Fuel Project (SNF Project) is responsible to DOE for operation of these fuel storage pools and for the 2100 metric tons of SNF materials that they contain. The SNF Project mission includes safe removal and transportation of all SNF from these storage basins to a new storage facility in the 200 East Area. To accomplish this mission, the SNF Project modifies the existing KE Basin and KW Basin facilities and constructs two new facilities: the 100 K Area Cold Vacuum Drying Facility (CVDF), which drains and dries the SNF; and the 200 East Area Canister Storage Building (CSB), which stores the SNF. The purpose of this document is to describe the design basis feed compositions for materials stored or processed by SNF Project facilities and activities. This document is not intended to replace the Hanford Spent Fuel Inventory Baseline (WHC 1994b), but only to supplement it by providing more detail on the chemical and radiological inventories in the fuel (this volume) and sludge. A variety of feed definitions is required to support evaluation of specific facility and process considerations during the development of these new facilities. Six separate feed types have been identified for development of new storage or processing facilities. The approach for using each feed during design evaluations is to calculate the proposed facility flowsheet assuming each feed. The process flowsheet would then provide a basis for material compositions and quantities which are used in follow-on calculations.

  20. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    SciTech Connect (OSTI)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  1. Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste Rob P. Rechard a,n , Geoff A. Freeze b , Frank V. Perry c a Nuclear Waste Disposal Research & Analysis, Sandia National Laboratories, P.O. Box 5800, Albuquerque 87185-0747, NM, USA b Applied Systems Analysis & Research, Sandia National Laboratories, P.O. Box 5800, Albuquerque 87185-0747, NM, USA c Earth and Environmental Sciences Division, Los Alamos National

  2. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect (OSTI)

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle; Wagner, John C.

    2013-07-01

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

  3. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports

    SciTech Connect (OSTI)

    Not Available

    1993-11-01

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

  4. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  5. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    SciTech Connect (OSTI)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  6. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    SciTech Connect (OSTI)

    Not Available

    1994-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  7. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 8

    SciTech Connect (OSTI)

    Payton, M. L.; Williams, J. T.; Tolbert-Smith, M.; Klein, J. A.

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  8. Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel. [Fuel crud

    SciTech Connect (OSTI)

    Hazelton, R.F.

    1987-09-01

    Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs.

  9. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices

    SciTech Connect (OSTI)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-10-01

    The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

  10. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    This document analyzes at a pregrammatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For pregrammatic spent nuclear fuel management, this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  11. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 2, Part A

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  12. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    SciTech Connect (OSTI)

    Rechard, R.P.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  13. A 2-D Test Problem for CFD Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields

    SciTech Connect (OSTI)

    Zigh, Ghani; Solis, Jorge; Fort, James A.

    2011-01-14

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 5-10 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper presents results for a simple 2-D problem that is an effective numerical analog for the neutron shield application. Because it is 2-D, solutions can be obtained relatively quickly allowing a comparison and assessment of sensitivity to model parameter changes. Turbulence models are considered as well as the tradeoff between steady state and transient solutions. Solutions are compared for two commercial CFD codes, FLUENT and STAR-CCM+. The results can be used to provide input to the CFD Best Practices for this application. Following study results for the 2-D test problem, a comparison of simulation results is provided for a high Rayleigh number experiment with large annular gap. Because the geometry of this validation is significantly different from the neutron shield, and due to the critical nature of this application, the argument is made for new experiments at representative scales

  14. Preliminary Microstructural Characterization of Gadolinium-Enriched Stainless Steels for Spent Nuclear Fuel Baskets (title change from A)

    SciTech Connect (OSTI)

    DUPONT,J.N.; ROBINO,CHARLES V.; STEPHENS JR.,JOHN J.; MCCONNELL,PAUL E.; MIZIA,R.; BRANAGAN,D.

    2000-07-24

    Gadolinium (Gd) is a very potent neutron absorber that can potentially provide the nuclear criticality safety required for interim storage, transport, and final disposal of spent nuclear fuel. Gd could be incorporated into an alloy that can be fabricated into baskets to provide structural support, corrosion resistance, and nuclear criticality control. In particular, Gd alloyed with stainless steel has been identified as a material that may fulfill these functional requirements. However, no information is available in the open literature that describes the influence of Gd on the microstructure and resultant mechanical properties of stainless steels alloyed with Gd. Such information is vital for determination of the suitability of these types of alloys for the intended application. Characterization of Gd-stainless steel (Gd-SS) alloys is also necessary for an American Society for Testing and Materials (ASTM) material specification, subsequent code approval by the American Society of Mechanical Engineers (ASME), and regulatory approval by the Nuclear Regulatory Commission for subsequent use by the nuclear industry. The Department of Energy National Spent Nuclear Fuel Program at Idaho National Engineering and Environmental Laboratory has commissioned Lehigh University and Sandia National Laboratories to characterize the properties of a series of Gd-SS alloys to assess their suitability for the spent fuel basket application. Preliminary microstructural characterization results are presented on Gd stainless steels. Small gas tungsten arc buttons were prepared by melting 316L stainless steel with 0.1 to 10 wt.% Gd. These samples were characterized by light optical and electron optical microscopy to determine the distribution of alloying elements and volume fraction of Gd-rich phase. The results acquired to date indicate that no Gd is dissolved in the austenite matrix. Instead, the Gd was present as an interdendritic constituent, and the amount of the Gd-rich constituent increased with nominal Gd concentration. The microstructure were similar to berated stainless steels in that each alloy system contains a hard secondary constituent dispersed in a ductile austenitic matrix. Microstructure-mechanical property correlations were therefore developed from previous work on berated stainless steels in order to guide selection of compositions of larger scale Gd-alloyed heats. In turn, these large-scale heats will form the basis for further investigations in which detailed microstructure, mechanical property, and corrosion resistance relationships will be developed.

  15. Room at the Mountain: Estimated Maximum Amounts of Commercial Spent Nuclear Fuel Capable of Disposal in a Yucca Mountain Repository

    SciTech Connect (OSTI)

    Kessler, John H. [Electric Power Research Institute - EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States); Kemeny, John [University of Arizona, Tucson AZ 85721 (United States); King, Fraser [Integrity Corrosion Consulting, Ltd., 6732 Silverview Drive NW, Calgary, Alberta (Canada); Ross, Alan M. [Alan M. Ross and Associates, 1061 Gray Fox Circle Pleasanton, CA 94566 (Canada); Ross, Benjamen [Disposal Safety, Inc., Bethesda, MD 20814 (United States)

    2006-07-01

    The purpose of this paper is to present an initial analysis of the maximum amount of commercial spent nuclear fuel (CSNF) that could be emplaced into a geological repository at Yucca Mountain. This analysis identifies and uses programmatic, material, and geological constraints and factors that affect this estimation of maximum amount of CSNF for disposal. The conclusion of this initial analysis is that the current legislative limit on Yucca Mountain disposal capacity, 63,000 MTHM of CSNF, is a small fraction of the available physical capacity of the Yucca Mountain system assuming the current high-temperature operating mode (HTOM) design. EPRI is confident that at least four times the legislative limit for CSNF ({approx}260,000 MTHM) can be emplaced in the Yucca Mountain system. It is possible that with additional site characterization, upwards of nine times the legislative limit ({approx}570,000 MTHM) could be emplaced. (authors)

  16. Public acceptability of the use of gamma rays from spent nuclear fuel as a hazardous waste treatment process

    SciTech Connect (OSTI)

    Mincher, B.J.; Wells, R.P.; Reilly, H.J.

    1992-01-01

    Three methods were used to estimate public reaction to the use of gamma irradiation of hazardous wastes as a hazardous waste treatment process. The gamma source of interest is spent nuclear fuel. The first method is Benefit-Risk Decision Making, where the benefits of the proposed technology are compared to its risks. The second analysis compares the proposed technology to the other, currently used nuclear technologies and estimates public reaction based on that comparison. The third analysis is called Analysis of Public Consent, and is based on the professional methods of the Institute for Participatory Management and Planning. The conclusion of all three methods is that the proposed technology should not result in negative public reaction sufficient to prevent implementation.

  17. Spent nuclear fuel project multi-year work plan WBS {number_sign}1.4.1

    SciTech Connect (OSTI)

    Wells, J.L.

    1997-03-01

    The Spent Nuclear Fuel (SNF) Project Multi-Year Work Plan (MYWP) is a controlled living document that contains the current SNF Project Technical, Schedule and Cost Baselines. These baselines reflect the current Project execution strategies and are controlled via the change control process. Other changes to the MYWP document will be controlled using the document control process. These changes will be processed as they are approved to keep the MYWP a living document. The MYWP will be maintained continuously as the project baseline through the life of the project and not revised annually. The MYWP is the one document which summarizes and links these three baselines in one place. Supporting documentation for each baseline referred to herein may be impacted by changes to the MYWP, and must also be revised through change control to maintain consistency.

  18. Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013

    SciTech Connect (OSTI)

    SISKIND B.; N /A

    2013-06-03

    This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

  19. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report Project A.5 and A.6

    SciTech Connect (OSTI)

    ARD, K.E.

    2000-04-19

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01.

  20. Study of charged particle motion in fields of different configurations for developing the concept of plasma separation of spent nuclear fuel

    SciTech Connect (OSTI)

    Smirnov, V. P.; Samokhin, A. A.; Vorona, N. A.; Gavrikov, A. V., E-mail: gavrikov@ihed.ras.ru [Russian Academy of Sciences, Joint Institute for High Temperatures (Russian Federation)

    2013-06-15

    The concept of plasma separation of spent nuclear fuel in a plane perpendicular to the magnetic field in an electric potential of special configuration is developed. A specific feature of the proposed approach consists in using an accelerating potential for reducing energy and angular spread of plasma ions at the entrance to the separator chamber and a potential well for the spatial separation of ions with different masses. The trajectories of ions of the substance imitating spent nuclear fuel are calculated. The calculations are performed for azimuthal and axial magnetic fields and model electric field configurations corresponding to different geometries of the separator chamber. It is shown that, using magnetic fields with a characteristic strength of 1 kG and electric potentials of up to 1 kV inside a region with a linear size less than 100 cm, it is possible to separate ions of spent nuclear fuel with energies from 0.2 to 3 eV. The calculations were performed for a collisionless mode in the single-particle approximation. Possible variants of the experimental facility for plasma separation of spent nuclear fuel are proposed.

  1. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents

    SciTech Connect (OSTI)

    Not Available

    1993-11-01

    The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

  2. Simultaneous recovery of all actinides from spent nuclear fuel by carbamoyl-methylphosphine oxide in fluorinated diluents

    SciTech Connect (OSTI)

    Ozawa, M.; Iwai, T.; Babain, V.; Shadrin, A.

    2008-07-01

    Bifunctional organophosphorus extractants dissolved in polar fluorinated diluents were studied, aiming at directly recovering all f-elements from the dissolver solution of spent nuclear fuel. Octyl(phenyl)-N,N-diisobutyl-carbamoyl-methylphosphine oxide (0{phi}D[iB] CMPO, 0.2-0.8 M) with 30% TBP dissolved in meta-nitrobenzotrifluoride (Fluoropole-732) dramatically expanded its extraction region without splitting out a heterogeneous third phase. Distribution ratios of U, Np, and Pu were sufficiently high for 0.4-0.8 M CMPO in this solvent system. Combination of salt-free, methylamine carbonate (MAC), citric acid, and hydrazine reagents were evaluated to obtain fractional stripping of f-elements such as TRU group and U. Static multistage extraction using artificial FBR dissolver solution supported the process feasibility. When all f-elements are extracted simultaneously, and TRU and U recovered separately with a single extraction cycle, the new extraction process, named ORGA-process, can be expected to be highly proliferation-resistant and systematically and economically advantageous. (authors)

  3. Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

    SciTech Connect (OSTI)

    L.M. Montierth

    2000-09-15

    The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issues and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).

  4. Transportation of Spent Nuclear Fuel and High Level Waste to Yucca Mountain: The Next Step in Nevada

    SciTech Connect (OSTI)

    Sweeney, Robin L,; Lechel, David J.

    2003-02-25

    In the U.S. Department of Energy's ''Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,'' the Department states that certain broad transportation-related decisions can be made. These include the choice of a mode of transportation nationally (mostly legal-weight truck or mostly rail) and in Nevada (mostly rail, mostly legal-weight truck, or mostly heavy-haul truck with use of an associated intermodal transfer station), as well as the choice among alternative rail corridors or heavy-haul truck routes with use of an associated intermodal transfer station in Nevada. Although a rail line does not service the Yucca Mountain site, the Department has identified mostly rail as its preferred mode of transportation, both nationally and in the State of Nevada. If mostly rail is selected for Nevada, the Department would then identify a preference for one of the rail corridors in consultation with affected stakeholders, particularly the State of Nevada. DOE would then select the rail corridor and initiate a process to select a specific rail alignment within the corridor for the construction of a rail line. Five proposed rail corridors were analyzed in the Final Environmental Impact Statement. The assessment considered the impacts of constructing a branch rail line in the five 400-meter (0.25mile) wide corridors. Each corridor connects the Yucca Mountain site with an existing mainline railroad in Nevada.

  5. Estimates of in situ deformability with an NX borehole jack, Spent Fuel Test - Climax, Nevada test site

    SciTech Connect (OSTI)

    Patrick, W.C.; Yow, J.L. Jr.; Axelrod, M.C.

    1985-12-01

    A series of borehole modulus measurements was obtained at the Spent Fuel Test - Climax (SFT-C) facility following removal of heat sources and a subsequent 1-year cooling period. A total of 212 measurements were obtained using a standard hardrock NX borehole (Goodman) jack. The results of 64 measurements made at the site before heating were reanalyzed for comparison with the post-heat data. Modulus values were calculated from the straight-line portion of the pressure vs displacement curves. Although the deformation modulus was observed to be highly variable, models were developed to explain much of this variability. Typically, spacial effects, anisotropy, and heating effects were present. The test results indicate that the deformation modulus tended to increase in the pillars between the underground openings where temperatures increased about 10{sup 0}C above the ambient 24{sup 0}C during the SFT-C. Conversely, a decrease in modulus was observed where temperatures were near 60{sup 0}C for a three-year period. In most cases, we found the modulus values to be slightly higher for vertical than for horizontal loading. There was a tendency for the modulus to be lower near excavated openings. While this effect was not ubiquitous, it was statistically significant.

  6. Maine Yankee: Making the Transition from an Operating Plant to an Independent Spent Fuel Storage Installation (ISFSI)

    SciTech Connect (OSTI)

    Norton, W.; McGough, M. S.

    2002-02-26

    The purpose of this paper is to describe the challenges faced by Maine Yankee Atomic Power Company in making the transition from an operating nuclear power plant to an Independent Spent Fuel Storage Installation (ISFSI). Maine Yankee (MY) is a 900-megawatt Combustion Engineering pressurized water reactor whose architect engineer was Stone & Webster. Maine Yankee was put into commercial operation on December 28, 1972. It is located on an 820-acre site, on the shores of the Back River in Wiscasset, Maine about 40 miles northeast of Portland, Maine. During its operating life, it generated about 1.2 billion kilowatts of power, providing 25% of Maine's electric power needs and serving additional customers in New England. Maine Yankee's lifetime capacity factor was about 67% and it employed more than 450 people. The decision was made to shutdown Maine Yankee in August of 1997, based on economic reasons. Once this decision was made planning began on how to accomplish safe and cost effective decommissioning of the plant by 2004 while being responsive to the community and employees.

  7. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California.

    SciTech Connect (OSTI)

    Bryan, Charles R.; Enos, David George

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  8. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    SciTech Connect (OSTI)

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called direct-use material which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  9. Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

  10. Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Dobbins, J.C.; Zaloudek, F.R.

    1987-07-01

    This report summarizes the histories of 17 Zircaloy-clad spent fuel assemblies used in dry storage tests and demonstrations at the Engine Maintenance and Disassembly (EMAD) and Climax facilities at the Nevada Test Site (NTS). The 18th assembly was shipped to the Battelle Columbus Laboratory (BCL) and remained there for extensive characterization and as a source of specimens for whole-rod and rod-segment dry storage tests. The report traces the history of the assemblies after discharge from the Turkey Point Unit 3 pressurized-water reactor (1975 and 1977) through shipment (first arrival at EMAD in December 1978), dry storage tests and demonstrations, and shipment by truck cask from EMAD to the Idaho National Engineering Laboratory (INEL) in May/June 1986. The principal objectives of this report are to assess and document the integrity of the fuel during the extensive dry storage activities at NTS and BCL, and to briefly summarize the dry storage technologies and procedures demonstrated in this program. The dry storage tests and demonstrations involved the following concepts and facilities: (1) surface drywells (EMAD); (2) deep drywells (425 m underground in the Climax granite formation); (3) concrete silo (EMAD); (4) air-cooled vault (EMAD); (5) electrically-heated module for fuel assembly thermal calibration and testing (EMAD/FAITM). 20 refs., 43 figs., 9 tabs.

  11. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Re, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  12. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect (OSTI)

    Bae, G.; Hong, S. G.

    2013-07-01

    In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

  13. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    SciTech Connect (OSTI)

    Rechard, R.P.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  14. Inspection and Gamma-Ray Dose Rate Measurements of the Annulus of the VSC-17 Concrete Spent Nuclear Fuel Storage Cask

    SciTech Connect (OSTI)

    P. L. Winston

    2007-09-01

    The air cooling annulus of the Ventilated Storage Cask (VSC)-17 spent fuel storage cask was inspected using a Toshiba 7 mm (1/4) CCD video camera. The dose rates observed in the annular space were measured to provide a reference for the activity to which the camera(s) being tested were being exposed. No gross degradation, pitting, or general corrosion was observed.

  15. Preliminary Simulations for Geometric Optimization of a High-Energy Delayed Gamma Spectrometer for Direct Assay of Pu in Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Kulisek, Jonathan A.; Campbell, Luke W.; Rodriguez, Douglas C.

    2012-06-07

    High-energy, beta-delayed gamma-ray spectroscopy is under investigation as part of the Next Generation Safeguard Initiative effort to develop non-destructive assay instruments for plutonium mass quantification in spent nuclear fuel assemblies. Results obtained to date indicate that individual isotope-specific signatures contained in the delayed gamma-ray spectra can potentially be used to quantify the total fissile content and individual weight fractions of fissile and fertile nuclides present in spent fuel. Adequate assay precision for inventory analysis can be obtained using a neutron generator of sufficient strength and currently available detection technology. In an attempt to optimize the geometric configuration and material composition for a delayed gamma measurement on spent fuel, the current study applies MCNPX, a Monte Carlo radiation transport code, in order to obtain the best signal-to-noise ratio. Results are presented for optimizing the neutron spectrum tailoring material, geometries to maximize thermal or fast fissions from a given neutron source, and detector location to allow an acceptable delayed gamma-ray signal while achieving a reasonable detector lifetime while operating in a high-energy neutron field. This work is supported in part by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  16. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement; Volume 1, Appendix F, Nevada Test Site and Oak Ridge Reservation Spent Nuclear Fuel Management Programs

    SciTech Connect (OSTI)

    1994-06-01

    This volume addresses the interim storage of spent nuclear fuel (SNF) at two US Department of Energy sites, the Nevada Test Site (NTS) and the Oak Ridge Reservation (ORR). These sites are being considered to provide a reasonable range of alternative settings at which future SNF management activities could be conducted. These locations are not currently involved in management of large quantities of SNF; NTS has none, and ORR has only small quantities. But NTS and ORR do offer experience and infrastructure for the handling, processing and storage of radioactive materials, and they do exemplify a broad spectrum of environmental parameters. This broad spectrum of environmental parameters will provide, a perspective on whether and how such location attributes may relate to potential environmental impacts. Consideration of these two sites will permit a programmatic decision to be based upon an assessment of the feasible options without bias, to the current storage sites. This volume is divided into four parts. Part One is the volume introduction. Part Two contains chapters one through five for the NTS, as well as references contained in chapter six. Part Three contains chapters one through five for the ORR, as well as references contained in chapter six. Part Four is summary information including the list of preparers, organizations contacted, acronyms, and abbreviations for both the NTS and the ORR. A Table of Contents, List of Figures, and List of Tables are included in parts Two, Three, and Four. This approach permitted the inclusion of both sites in one volume while maintaining consistent chapter numbering.

  17. Assessment of effectiveness of geologic isolation systems. Test case release consequence analysis for a spent fuel repository in bedded salt

    SciTech Connect (OSTI)

    Raymond, J.R.; Bond, F.W.; Cole, C.R.; Nelson, R.W.; Reisenauer, A.E.; Washburn, J.F.; Norman, N.A.; Mote, P.A.; Segol, G.

    1980-01-01

    Geologic and geohydrologic data for the Paradox Basin have been used to simulate movement of ground water and radioacrtive contaminants from a hypothetical nuclear reactor spent fuel repository after an assumed accidental release. The pathlines, travel times and velocity of the ground water from the repository to the discharge locale (river) were determined after the disruptive event by use of a two-dimensional finite difference hydrologic model. The concentration of radioactive contaminants in the ground water was calculated along a series of flow tubes by use of a one-dimensional mass transport model which takes into account convection, dispersion, contaminant/media interactions and radioactive decay. For the hypothetical site location and specific parameters used in this demonstration, it is found that Iodine-129 (I-129) is tthe only isotope reaching the Colorado River in significant concentration. This concentration occurs about 8.0 x 10/sup 5/ years after the repository has been breached. This I-129 ground-water concentration is about 0.3 of the drinking water standard for uncontrolled use. The groundwater concentration would then be diluted by the Colorado River. None of the actinide elements reach more than half the distance from the repository to the Colorado River in the two-million year model run time. This exercise demonstrates that the WISAP model system is applicable for analysis of contaminant transport. The results presented in this report, however, are valid only for one particular set of parameters. A complete sensitivity analysis must be performed to evaluate the range of effects from the release of contaminants from a breached repository.

  18. A comparison of surface topography characterization technologies for use in comparing spent bullet and cartridge case signatures

    SciTech Connect (OSTI)

    Batishko, C.R.; Hickman, B.J.; Cuta, F.M.

    1992-11-01

    The Pacific Northwest Laboratory was tasked by the US Department of Energy to provide technical assistance to the Federal Bureau of Investigation in evaluating and ranking technologies potentially useful in high-speed comparison of unique spent bullet and cartridge case surface signatures. Information sources included vendor input, current relevant literature, vendor phone contacts, other FBI resources, relevant PNL reports, and personal contact with numerous PNL technical staff. A comprehensive list of technologies was reduced to a list of 38 by grouping very similar methodologies, and further reduced to a short list of six by applying a set of five minimum functional requirements. A total of 14 primary criteria, many having secondary criteria, were subsequently used to evaluate each technology. The ranked short list results are reported and supported in this document, and their scores normalized to a hypothetical ideal system are as follows: (1) confocal microscopy 82.13; (2) laser dynamic focusing 72.04; (3)moire interferometry V70.94; (4)fringe field capacitance;(5)laser triangulation 66.18; (6)structured/sectioned light 65.55. Information available within the time/budget constraints which was used for the evaluation and ranking was not sufficiently detailed to evaluate specific implementations of the technologies. Each of the technologies in the short list was judged potentially capable of meeting the minimum requirements. Clever, novel engineering solutions resulting in a more cost-effective system, or a closer fit to the ``ideal system,`` could result in a reordering of the short list when actual technical proposals are evaluated. Therefore, it is recommended that a Request for Proposal not be limited to only the highest ranked technology, but include all six technologies in the short list.

  19. A comparison of surface topography characterization technologies for use in comparing spent bullet and cartridge case signatures

    SciTech Connect (OSTI)

    Batishko, C.R.; Hickman, B.J.; Cuta, F.M.

    1992-11-01

    The Pacific Northwest Laboratory was tasked by the US Department of Energy to provide technical assistance to the Federal Bureau of Investigation in evaluating and ranking technologies potentially useful in high-speed comparison of unique spent bullet and cartridge case surface signatures. Information sources included vendor input, current relevant literature, vendor phone contacts, other FBI resources, relevant PNL reports, and personal contact with numerous PNL technical staff. A comprehensive list of technologies was reduced to a list of 38 by grouping very similar methodologies, and further reduced to a short list of six by applying a set of five minimum functional requirements. A total of 14 primary criteria, many having secondary criteria, were subsequently used to evaluate each technology. The ranked short list results are reported and supported in this document, and their scores normalized to a hypothetical ideal system are as follows: (1) confocal microscopy 82.13; (2) laser dynamic focusing 72.04; (3)moire interferometry V70.94; (4)fringe field capacitance;(5)laser triangulation 66.18; (6)structured/sectioned light 65.55. Information available within the time/budget constraints which was used for the evaluation and ranking was not sufficiently detailed to evaluate specific implementations of the technologies. Each of the technologies in the short list was judged potentially capable of meeting the minimum requirements. Clever, novel engineering solutions resulting in a more cost-effective system, or a closer fit to the ideal system,'' could result in a reordering of the short list when actual technical proposals are evaluated. Therefore, it is recommended that a Request for Proposal not be limited to only the highest ranked technology, but include all six technologies in the short list.

  20. Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage

    SciTech Connect (OSTI)

    Guenther, R.J.; Gilbert, E.R.; Slate, S.C.; Partain, W.L.; Divine, J.R.; Kreid, D.K.

    1984-01-01

    It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanisms have been identified that could, under certain conditions, cause failure of the SF cladding and/or failure of TRUW storage containers. Stress rupture limits for Zircaloy-clad SF in MRS range from 300 to 440/sup 0/C, based on limited data. Additional tests on irradiated Zircaloy (3- to 5-year duration) are needed to narrow this uncertainty. Cladding defect sizes could increase in air as a result of fuel density decreases due to oxidation. Oxidation tests (3- to 5-year duration) on SF are also needed to verify oxidation rates in air and to determine temperatures below which monitoring of an inert cover gas would not be required. Few, if any, changes in the physical state of HLW glass or canisters or their performance would occur under projected MRS conditions. The major uncertainty for HLW is in the heat transfer through cracked glass and glass devitrification above 500/sup 0/C. Additional study of TRUW is required. Some fraction of present TRUW containers would probably fail within the first 100 years of MRS, and some TRUW would be highly degraded upon retrieval, even in unfailed containers. One possible solution is the design of a 100-year container. 93 references, 28 figures, 17 tables.