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1

Pilot scale gasification of spent cooking liquor from sodium sulfite based delignification  

Science Journals Connector (OSTI)

This paper describes a pilot scale high pressure entrained flow gasification experiment with spent cooking liquor from a sodium sulfite based delignification process in the DP-1 black liquor gasifier in Piteå, Sweden. Approximately 92 tons of sulfite ...

Erik Furusjö; Ragnar Stare; Ingvar Landälv; Patrik Löwnertz

2014-11-17T23:59:59.000Z

2

Reductive burning of high-yield spent pulping liquors by the addition of pulverized coal  

SciTech Connect

This paper reports on the reductive burning of high-yield spent pulping liquors which can be accomplished by the addition of pulverized coal to increase the heat content and generate the proper reducing conditions. Samples from a 78%-yield sodium bisulfite pulping process employing a hardwood furnish were mixed with 10-50% pulveriized coal and burned at 950[degrees]C under reducing conditions in a box furnace. Even in these uncontrolled combustion conditions 76. 5% of the sulfur found in the soluble portion of the smelt was converted from lignousulfonates to useful sulfide ion. For the remainder of the sulfur, analyses determined it to be 19. 5% as sulfite ion, 3. 1% as thiosulfate ion, and 0.9% as sulfate ion.

Sell, N.J.; Norman, J.C. (Natural and Applied Sciences, Univ. of Wisconsin-Green Bay, Green Bay, WI (United States))

1992-10-01T23:59:59.000Z

3

Reducing NO[sub x] emissions from magnesium sulfite liquor boilers  

SciTech Connect

The Current design of liquor-combustion boilers is reviewed, along with options for lowering exhaust-gas emissions, particularly NOx. In many cases, modern boilers are being operated at or near optimal conditions for minimum NOx emissions. Possible upgrades to further reduce NOx emissions include addition of a selective noncatalytic reduction step, design modifications to lower the sectional load, a flue-gas recirculation system, and air staging. Calculated and experimental results show that these applications can lower NOx emissions by 40% or more.

Bobik, M. (Austrian Energy and Environment, Graz (Austria))

1993-01-01T23:59:59.000Z

4

Study on Conditioning of SO2–Ethanol–Water Spent Liquor from Spruce Chips/Softwood Biomass for ABE Fermentation  

Science Journals Connector (OSTI)

This study is an integral part of a project which aims at creating an economic process that can utilize cheap forestry residues such as twigs, cones, treetops, branches, and bark to produce renewable chemicals and liquid fuels. ... Further instruments were used for various analyses: furfural and hydroxymethylfurfural by HPLC (Dionex UltiMate 3000 equipped with diode array detector); acetic acid, formic acid, and ethanol by an HP 1100 HPLC; aldonic acids by HPAEC equipped with a Dionex CarboPac PA10 column; uronic acids by methanolysis/GC-FID (Shimadzu GC-2010 Plus with NB-30 capillary column of 30 m length and 0.32 mm interior diameter) according to Sundberg et al.(17) and Iakovlev and van Heiningen;(2) concentrations of sulfate and sulfite anions (after adding NaOH) by ion chromatography (Dionex ICS 1500, Sunnyvale, CA, USA); inorganic analysis of the feedstocks by a Varian Liberty ICP-AAS; CHN/S analysis of the feedstocks by a 2400 Series II CHN/S PerkinElmer elemental analyzer (Jones factor of 6.25 used to convert nitrogen content to protein content according to Mariotti et al.(18)); sulfur content in the LCC by combustion in oxygen in a Schöniger flask, followed by absorption of SO2 in H2O2 solution, and the formed sulfate anions were determined by ion chromatography (SCAN-CM 57:99). ...

Evangelos Sklavounos; Mikhail Iakovlev; Adriaan van Heiningen

2013-02-26T23:59:59.000Z

5

The preparation and analysis of ammonia base sulfite pulping liquor  

E-Print Network (OSTI)

NH3 Concentrations . . . . . , ~ . ~ . . ~ ~ . ~ q IV The Distribution of S02 in a 2. 82$ RR3 Solution for Various Total SO2 Concentrations . . . . . . , . . . ~ . . . 31 FIGURES l. Variation of pH with Percent NH3 in Solution. . . . . ~. . . 25... 2. Variation of pH with S02 Concentration in Solutions of Various NH Concentrations . . . . ~ . . . ~ . ~ o ~ . o . ~ 27 3 ~ Composition of Solutions to give a Constant pH of T and 2. 5 , ~ 2Q 4. The Variation of the Ratio of Free S02 to Total SO2...

Honstead, John Frederick

1950-01-01T23:59:59.000Z

6

Kraft Liquor Corrosion Margaret Gorog  

E-Print Network (OSTI)

1 Kraft Liquor Corrosion Margaret Gorog Federal Way, WA Pulp and Paper Corrosion Symposium Georgia Tech Renewable Bioproducts Institute November 2014 · Brown Stock Corrosion · Alkaline Liquor Corrosion · Black Liquor Corrosion ­ Evaporators ­ Research · High Solids Black Liquor Corrosion of Stainless Steel

Das, Suman

7

Chapter 2 - Black Liquor Gasification  

Science Journals Connector (OSTI)

Black liquor gasification (BLG) is being considered primarily as an option for production of biofuels in recent years due to the focus on the transport sector’s high oil dependence and climate impact. BLG may be performed either at low temperatures or at high temperatures, based on whether the process is conducted above or below the melting temperature range of the spent pulping chemicals. The development of various BLG technologies—SCA-Billerud process, the Copeland recovery process, Weyerhaeuser’s process, the St. Regis hydropyrolysis process, the Texaco process, VTT’s circulating fluidized bed BLG process, Babcock and Wilcox’s bubbling fluidized bed gasification process, NSP process (Ny Sodahus Process), DARS (Direct Alkali Recovery System) process, BLG with direct causticization, Manufacturing and Technology Conversion International fluidized bed gasification, Chemrec gasification, catalytic hydrothermal gasification of black liquor—is discussed in this chapter. The two main technologies under development are pressurized gasification and atmospheric gasification, being commercialized by Chemrec AB and ThermoChem Recovery International, respectively.

Pratima Bajpai

2014-01-01T23:59:59.000Z

8

Gasification of black liquor  

DOE Patents (OSTI)

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.

Kohl, Arthur L. (Woodland Hills, CA)

1987-07-28T23:59:59.000Z

9

Proceedings of the black liquor research program review fifth meeting  

SciTech Connect

On June 14--17, 1988 the participants and invited guests of the Cooperative Program in Kraft Recovery gathered in Charleston, South Carolina, to review progress on four major black liquor research programs being executed at the Institute of Paper Chemistry, the University of Maine, the National Bureau of Standards, and the University of Florida. These programs include: (1) Black Liquor Properties; (2) Black Liquor Droplet Formation; (3) Black Liquor Nozzle Evaluation; and (4) Black Liquor Combustion. In addition to the objectives of previous meetings, this meeting made a direct attempt to gather ideas on how to improve our ability to move from new technology concepts to commercial implementation. Also attached is the agenda for the Charleston meeting. The first two days were involved with updates and reviews of the four major black liquor programs. A half day was spent discussing pathways to implementation and developing thoughts on what industry, DOE and academia could do to facilitate commercial implementation of the research results. This publication is a summary of the presentations made in Charleston and the industry responses to the research work. Readers are cautioned that the contents are in-progress updates on the status of the research and do not represent referred technical papers. Any questions regarding the content should be referred to the principal investigators of the project.

Not Available

1988-09-01T23:59:59.000Z

10

CORN STEEP LIQUOR IN MICROBIOLOGY  

Science Journals Connector (OSTI)

...is a by-product of the corn wet-milling industry it would be insufficient...invention of much of the modern wet milling process, suggested corn steep liquor as a nutrient...general flowsheet of the corn wet-milling process; and to Dr. L...

R. Winston Liggett; H. Koffler

1948-12-01T23:59:59.000Z

11

NETL: News Release - DOE Opens Competition for Black Liquor/Biomass  

NLE Websites -- All DOE Office Websites (Extended Search)

January 7, 2000 January 7, 2000 DOE Opens Competition for Black Liquor/Biomass Gasification Program Intended to Boost Efficiency, Reduce Greenhouse Gases from Pulp and Paper Mills A new competition begun this week by the Department of Energy could make the pulp and paper mills of the 21st century cleaner and more energy efficient by demonstrating improved technologies to convert their spent cooking liquor streams into new sources of energy. The advanced processes would also simultaneously recover and recycle pulping chemicals. MORE INFO Download the solicitation The department's National Energy Technology Laboratory has issued a call for projects to demonstrate advanced ways to gasify the black liquor or biomass of pulp and paper mills. The gases can be more easily cleaned of

12

NETL: Utilization Projects - Value Added Products from FGD Sulfite rich  

NLE Websites -- All DOE Office Websites (Extended Search)

Value Added Products from FGD Sulfite rich Scrubber Material Value Added Products from FGD Sulfite rich Scrubber Material In pursuit of developing value added products from sulfite-rich scrubber material, e.g., low-density panels, carpet underlayment, siding, pre-cast building material, lumber panels, particle and wafer type boards, the following four experimental tasks are proposed: A comprehensive characterization of sulfite-rich scrubber materials produced by power plant generation. Specifically, the mercury, selenium, arsenic, boron, and organic content will be monitored The sulfite-rich scrubber material will be combined with cheap but renewable agricultural byproducts like micronized core fibers and/or micronized wheat straw, and the composites will be formulated by exploiting the natural polymers of the byproducts. The conditions under which structural composites can be formulated using injection molding and compressive molding will be evaluated.

13

Fuel nitrogen release during black liquor pyrolysis; Part 2: Comparisons between different liquors  

SciTech Connect

This continuation of earlier work reports fuel nitrogen release for black liquors at two temperatures during pyrolysis of single droplets in an oxygen-free environment. Approximately half of the 20--60% fuel nitrogen released was ammonia and half was molecular nitrogen. The total amount of fixed nitrogen released during pyrolysis was almost linearly proportional to the liquor nitrogen content. The yield of fixed nitrogen for birch liquors was significantly higher than for pine liquors, and the yield for bagasse liquor was extremely high.

Aho, K.; Nikkanen, S. (A. Ahlstrom Corp., Varkaus (Finland)); Hupa, M. (Abo Akademi Univ., Turku (Finland). Chemical Engineering Dept.)

1994-08-01T23:59:59.000Z

14

Highly Energy Efficient Directed Green Liquor Utilization (D...  

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

Highly Energy Efficient Directed Green Liquor Utilization (D-GLU) Pulping Highly Energy Efficient Directed Green Liquor Utilization (D-GLU) Pulping This factsheet describes a...

15

Refractory for Black Liquor Gasifiers  

SciTech Connect

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.

William L. Headrick Jr; Alireza Rezaie

2003-12-01T23:59:59.000Z

16

Refractory for Black Liquor Gasifiers  

SciTech Connect

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.

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

2005-12-01T23:59:59.000Z

17

REFRACTORY FOR BLACK LIQUOR GASIFIERS  

SciTech Connect

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.

William L. Headrick Jr.; Alireza Rezaie

2003-12-01T23:59:59.000Z

18

Refractory for Black Liquor Gasifiers  

SciTech Connect

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.

William L. Headrick Jr; Alireza Rezaie

2003-08-01T23:59:59.000Z

19

Refractory for Black Liquor Gasifiers  

SciTech Connect

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.

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

2003-03-31T23:59:59.000Z

20

Mill Integration-Pulping, Stream Reforming and Direct Causticization for Black Liquor Recovery  

SciTech Connect

MTCI/StoneChem developed a steam reforming, fluidized bed gasification technology for biomass. DOE supported the demonstration of this technology for gasification of spent wood pulping liquor (or 'black liquor') at Georgia-Pacific's Big Island, Virginia mill. The present pre-commercial R&D project addressed the opportunities as well as identified negative aspects when the MTCI/StoneChem gasification technology is integrated in a pulp mill production facility. The opportunities arise because black liquor gasification produces sulfur (as H{sub 2}S) and sodium (as Na{sub 2}CO{sub 3}) in separate streams which may be used beneficially for improved pulp yield and properties. The negative aspect of kraft black liquor gasification is that the amount of Na{sub 2}CO{sub 3} which must be converted to NaOH (the so called causticizing requirement) is increased. This arises because sulfur is released as Na{sub 2}S during conventional kraft black liquor recovery, while during gasification the sodium associated Na{sub 2}S is partly or fully converted to Na{sub 2}CO{sub 3}. The causticizing requirement can be eliminated by including a TiO{sub 2} based cyclic process called direct causticization. In this process black liquor is gasified in the presence of (low sodium content) titanates which convert Na{sub 2}CO{sub 3} to (high sodium content) titanates. NaOH is formed when contacting the latter titanates with water, thereby eliminating the causticizing requirement entirely. The leached and low sodium titanates are returned to the gasification process. The project team comprised the University of Maine (UM), North Carolina State University (NCSU) and MTCI/ThermoChem. NCSU and MTCI are subcontractors to UM. The principal organization for the contract is UM. NCSU investigated the techno-economics of using advanced pulping techniques which fully utilize the unique cooking liquors produced by steam reforming of black liquor (Task 1). UM studied the kinetics and agglomeration problems of the conversion of Na{sub 2}CO{sub 3} to (high sodium) titanates during gasification of black liquor in the presence of (low sodium) titanates or TiO{sub 2} (Task 2). MTCI/ThermoChem tested the performance and operability of the combined technology of steam reforming and direct causticization in their Process Development Unit (PDU) (Task 3). The specific objectives were: (1) to investigate how split sulfidity and polysulfide (+ AQ) pulping can be used to increase pulp fiber yield and properties compared to conventional kraft pulping; (2) to determine the economics of black liquor gasification combined with these pulping technologies in comparison with conventional kraft pulping and black liquor recovery; (3) to determine the effect of operating conditions on the kinetics of the titanate-based direct causticization reaction during black liquor gasification at relatively low temperatures ({le} 750 C); (4) to determine the mechanism of particle agglomeration during gasification of black liquor in the presence of titanates at relatively low temperatures ({le} 750 C); and (5) to verify performance and operability of the combined technology of steam reforming and direct causticization of black liquor in a pilot scale fluidized bed test facility.

Adriaan van Heiningen

2007-06-30T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Recovering Soaps and Greases from Wool-Scouring liquors  

Science Journals Connector (OSTI)

Recovering Soaps and Greases from Wool-Scouring liquors ... A NEW process, illustrated in the accompanying flowsheet, has been devised for recovering soaps and greases from waste wool-scouring liquors. ...

C.H.S. TUPHOLME

1939-12-20T23:59:59.000Z

22

Demonstration of Black Liquor Gasification at Big Island  

SciTech Connect

This Final Technical Report provides an account of the project for the demonstration of Black Liquor Gasification at Georgia-Pacific LLC's Big Island, VA facility. This report covers the period from May 5, 2000 through November 30, 2006.

Robert DeCarrera

2007-04-14T23:59:59.000Z

23

Bench-scale Kinetics Study of Mercury Reactions in FGD Liquors  

SciTech Connect

This document is the final report for Cooperative Agreement DE-FC26-04NT42314, 'Kinetics Study of Mercury Reactions in FGD Liquors'. The project was co-funded by the U.S. DOE National Energy Technology Laboratory and EPRI. The objective of the project has been to determine the mechanisms and kinetics of the aqueous reactions of mercury absorbed by wet flue gas desulfurization (FGD) systems, and develop a kinetics model to predict mercury reactions in wet FGD systems. The model may be used to determine optimum wet FGD design and operating conditions to maximize mercury capture in wet FGD systems. Initially, a series of bench-top, liquid-phase reactor tests were conducted and mercury species concentrations were measured by UV/visible light spectroscopy to determine reactant and byproduct concentrations over time. Other measurement methods, such as atomic absorption, were used to measure concentrations of vapor-phase elemental mercury, that cannot be measured by UV/visible light spectroscopy. Next, a series of bench-scale wet FGD simulation tests were conducted. Because of the significant effects of sulfite concentration on mercury re-emission rates, new methods were developed for operating and controlling the bench-scale FGD experiments. Approximately 140 bench-scale wet FGD tests were conducted and several unusual and pertinent effects of process chemistry on mercury re-emissions were identified and characterized. These data have been used to develop an empirically adjusted, theoretically based kinetics model to predict mercury species reactions in wet FGD systems. The model has been verified in tests conducted with the bench-scale wet FGD system, where both gas-phase and liquid-phase mercury concentrations were measured to determine if the model accurately predicts the tendency for mercury re-emissions. This report presents and discusses results from the initial laboratory kinetics measurements, the bench-scale wet FGD tests, and the kinetics modeling efforts.

Gary Blythe; John Currie; David DeBerry

2008-03-31T23:59:59.000Z

24

Value-Added Products from FGD Sulfite-Rich Scrubber Materials  

SciTech Connect

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.

Vivak Malhotra

2010-01-31T23:59:59.000Z

25

CHAMPAIGN-URBANA FOOD AND LIQUOR GUIDE 1. Introduction  

E-Print Network (OSTI)

(217) 352-7467. Excellent Mexican and Latin American- influenced food. Service can be a bit slowCHAMPAIGN-URBANA FOOD AND LIQUOR GUIDE 1. Introduction This is a short, incomplete, and personally. It is standard to tip 15-20% for reasonable service. A indicates a place that we like, and indicates a place we

Athreya, Jayadev

26

Report on the WORKSHOP ON COMMERCIALIZATION OF BLACK LIQUOR  

E-Print Network (OSTI)

!iers have not seen convIncIng evidence of cycle (BLGCC) and biomass-gasifier/combined market opportunity -Denny Hunter gasification/combined cycle technology, sufficient Weyerhaeuser Company research exists for Key ideas from the workshop... commercializing black liquor gasifier/ combined Techno~og~ supp

27

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

SciTech Connect

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.

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

1998-05-01T23:59:59.000Z

28

Conceptual design of a black liquor gasification pilot plant  

SciTech Connect

In July 1985, Champion International completed a study of kraft black liquor gasification and use of the product gases in a combined cycle cogeneration system based on gas turbines. That study indicated that gasification had high potential as an alternative to recovery boiler technology and offered many advantages. This paper describes the design of the plant, the construction of the pilot plant, and finally presents data from operation of the plant.

Kelleher, E. G.

1987-08-01T23:59:59.000Z

29

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

SciTech Connect

A wide variety of experimental techniques have been used in this work, and many of these have been developed completely or improved significantly in the course of the research done during this program. Therefore, it is appropriate to describe these techniques in detail as a reference for future workers so that the techniques can be used in future work with little additional effort or so that the results reported from this program can be compared better with future results from other work. In many cases, the techniques described are for specific analytical instruments. It is recognized that these may be superseded by future developments and improvements in instrumentation if a complete description of techniques used successfully in the past on other instrumentation is available. The total pulping and liquor preparation research work performed included chip and white liquor preparation, digestion, pulp washing, liquor and wash recovery, liquor sampling, weak liquor concentration in two steps to about 45--50% solids with an intermediate soap skimming at about 140F and 27--30% solids, determination of pulp yield and Kappa number, determination of total liquor solids, and a check on the total material balance for pulping. All other research was performed either on a sample of the weak black liquor (the combined black liquor and washes from the digester) or on the skimmed liquor that had been concentrated.

Fricke, A.L.; Zaman, A.A.; Stoy, M.O.; Schmidl, G.W.; Dong, D.J.; Speck, B.

1998-04-01T23:59:59.000Z

30

Chapter 4 - Obstacles to Implementation of Black Liquor Gasification  

Science Journals Connector (OSTI)

Obstacles hindering the commercialization of black liquor gasification are discussed. The most important of them include financial risks, protection of the gasifier, increase in the causticizing demand, tar condensation, hot gas cleanup, and steam deficit. Gasification system demands significant capital investment. The high temperature and pressure and the alkaline conditions create aggressive environment. Protection from an aggressive environment is very important in the operation of a gasifier because it determines the lifetime and hence the feasibility of gasification as a technology as a whole.

Pratima Bajpai

2014-01-01T23:59:59.000Z

31

Associations between Beer, Wine, and Liquor Consumption and Lung Cancer Risk: A Meta-analysis  

Science Journals Connector (OSTI)

...literature on beer, wine, and liquor intake and...alcoholic beverage (beer, wine, or liquor) and lung...per bottle, 150 mL of wine per glass, and 40 mL...beverage could potentially increase the risk of lung cancer...alcoholic beverages, and quality of smoking adjustment...

Chun Chao

2007-11-01T23:59:59.000Z

32

Chapter 5 - Environmental Impact of Black Liquor Gasification  

Science Journals Connector (OSTI)

Environmental impact of black liquor gasification (BLG) is discussed. Biofuels from a BLG process excel in terms of well-to-wheel carbon dioxide emission reduction and energy efficiency. Forest biorefinery utilizing gasification (in a black liquor gasification combined cycle (BLGCC) configuration) rather than a Tomlinson boiler is predicted to produce significantly fewer pollutant emissions due to the intrinsic characteristics of the BLGCC technology. Syngas cleanup conditioning removes a considerable amount of contaminants and gas turbine combustion is more efficient and complete than boiler combustion. Also, there could be reductions in pollutant emissions and hazardous wastes resulting from cleaner production of chemicals and fuels that are now manufactured using fossil energy resources. Production of power, fuels, chemicals, and other products from biomass resources creates a net zero generation of carbon dioxide as plants are renewable carbon sinks. BLG whether conducted at high or low temperatures is still superior to the current recovery boiler combustion technology. Implementation of IGCC power plants will cause net savings in cooling water requirements and net reductions in wastewater discharges. The most significant environmental impact caused by BLG will occur in air emissions. The overall reduction of Total reduced sulphur (TRS) gases using gasification technology will also reduce odor, which will improve public acceptance of pulp and paper mills, particularly in populated areas.

Pratima Bajpai

2014-01-01T23:59:59.000Z

33

MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University of Tennessee Georgia Tech SURF 2010 Fellow  

E-Print Network (OSTI)

MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University Goodman Introduction In the papermaking industry, black liquor recovery boilers burn black liquor into the superheater region of the boiler, where the salt-deposit, or smelt, forms a scale on the superheater tubes.1

Li, Mo

34

Biomass gasification project gets funding to solve black liquor safety and landfill problems  

SciTech Connect

This paper reports on biomass gasifications. The main by-product in pulp making is black liquor from virgin fiber; the main by-product in paper recycling is fiber residue. Although the black liquor is recycled for chemical and energy recovery, safety problems plague the boilers currently used to do this. The fiber residue is usually transported to a landfill. The system being developed by MTCI will convert black liquor and fiber residue into a combustible gas, which can then be used for a wide variety of thermal or power generation applications.

Black, N.P.

1991-02-01T23:59:59.000Z

35

Spent fuel pyroprocessing demonstration  

SciTech Connect

A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option.

McFarlane, L.F.; Lineberry, M.J.

1995-05-01T23:59:59.000Z

36

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

DOE Patents (OSTI)

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.

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

2008-06-24T23:59:59.000Z

37

Nuclear Spent Fuel Program Drivers  

NLE Websites -- All DOE Office Websites (Extended Search)

was created to plan and coordinate the management of Department of Energy-owned spent nuclear fuel. It was established as a result of a 1992 decision to stop spent nuclear fuel...

38

Fuel nitrogen release during black liquor pyrolysis; Part 1: Laboratory measurements at different conditions  

SciTech Connect

Fuel nitrogen release during black liquor pyrolysis is high. There is only minor release during the drying stage. Ammonia is the main fixed nitrogen species formed. The rate of fixed nitrogen release increases with increasing temperature. The level of fixed nitrogen released by birch liquor is almost twice the level for pine liquor. Assuming complete conversion to NO, fixed nitrogen yields gave NO concentrations near typically measured values for flue gases in full scale recovery boilers. The purpose of this work was to gain more detailed information about the behavior of the fuel nitrogen in black liquor combustion. The work focused on the pyrolysis or devolatilization of the combustion process. Devolatilization is the stage at which the majority (typically 50--80%) of the liquor organics release from a fuel particle or droplet as gaseous species due to the rapid destruction of the organic macromolecules in the liquor. In this paper, the authors use the terms devolatilization and pyrolysis interchangeably with no difference in their meaning.

Aho, K.; Vakkilainen, E. (A. Ahistrom Corp., Varkaus (Finland)); Hupa, M. (Abo Akademi Univ., Turku (Finland). Chemical Engineering Dept.)

1994-05-01T23:59:59.000Z

39

Improved Materials for High-Temperature Black Liquor Gasification  

SciTech Connect

The laboratory immersion test system built and operated at ORNL was found to successfully screen samples from numerous refractory suppliers, including both commercially available and experimental materials. This system was found to provide an accurate prediction of how these materials would perform in the actual gasifier environment. Test materials included mullites, alumino-silicate bricks, fusion-cast aluminas, alumina-based and chrome-containing mortars, phosphate-bonded mortars, coated samples provided under an MPLUS-funded project, bonded spinels, different fusion-cast magnesia-alumina spinels with magnesia content ranging from 2.5% to about 60%, high-MgO castable and brick materials, spinel castables, and alkali-aluminate materials. This testing identified several candidate material systems that perform well in the New Bern gasifier. Fusion-cast aluminas were found to survive for nearly one year, and magnesia-alumina spinels have operated successfully for 18 months and are expected to survive for two years. Alkali-aluminates and high-MgO-content materials have also been identified for backup lining applications. No other material with a similar structure and chemical composition to that of the fusion-cast magnesium-aluminum spinel brick currently being used for the hot-face lining is commercially available. Other materials used for this application have been found to have inferior service lives, as previously discussed. Further, over 100 laboratory immersion tests have been performed on other materials (both commercial and experimental), but none to date has performed as well as the material currently being used for the hot-face lining. Operating experience accumulated with the high-temperature gasifier at New Bern, North Carolina, has confirmed that the molten alkali salts degrade many types of refractories. Fusion-cast alumina materials were shown to provide a great improvement in lifetime over materials used previously. Further improvement was realized with fusion-cast magnesia-alumina spinel refractory, which appears to be the most resistant to degradation found to date, exhibiting over a year of service life and expected to be capable of over two years of service life. Regarding the use of refractory mortar, it was found that expansion of the current chrome-alumina mortar when subjected to black liquor smelt is likely contributing to the strains seen on the vessel shell. Additionally, the candidate high-alumina mortar that was originally proposed as a replacement for the current chrome-alumina mortar also showed a large amount of expansion when subjected to molten smelt. A UMR experimental mortar, composed of a phosphate bonded system specifically designed for use with fusion-cast magnesium-aluminum spinel, was found to perform well in the molten smelt environment. Strain gauges installed on the gasifier vessel shell provided valuable information about the expansion of the refractory, and a new set of strain gauges and thermocouples has been installed in order to monitor the loading caused by the currently installed spinel refractory. These results provide information for a direct comparison of the expansion of the two refractories. Measurements to date suggest that the fusion-cast magnesia-alumina spinel is expanding less than the fusion-cast {alpha}/{beta}-alumina used previously. A modified liquor nozzle was designed and constructed to test a number of materials that should be more resistant to erosion and corrosion than the material currently used. Inserts made of three erosion-resistant metallic materials were fabricated, along with inserts made of three ceramic materials. The assembled system was sent to the New Bern mill for installation in the gasifer in 2005. Following operation of the gasifier using the modified nozzle, inserts should be removed and analyzed for wear by erosion/corrosion. Although no materials have been directly identified for sensor/thermocouple protection tubes, several of the refractory material systems identified for lining material applications may be applicable for use in this

Keiser, J.R.; Hemrick, J.G.; Gorog, J.P.; Leary, R.

2006-06-29T23:59:59.000Z

40

Spent Nuclear Fuel Fact Sheets  

NLE Websites -- All DOE Office Websites (Extended Search)

management needs. By coordinating common needs for research, technology development, and testing programs, the National Spent Nuclear Fuel Program is achieving cost efficiencies...

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Microorganisms with Novel Dissimilatory (Bi)Sulfite Reductase Genes Are Widespread and Part of the Core Microbiota in Low-Sulfate Peatlands  

Science Journals Connector (OSTI)

...accession numbers) qPCR performance Mean efficiency () SD Linearity R 2 Dynamic range (no...cyanobacterial mats of Solar Lake (Sinai, Egypt). Appl. Environ. Microbiol. 64...sulfite, or some organosulfonates for energy conservation but can also be present in...

Doris Steger; Cecilia Wentrup; Christina Braunegger; Pinsurang Deevong; Manuel Hofer; Andreas Richter; Christian Baranyi; Michael Pester; Michael Wagner; Alexander Loy

2010-12-17T23:59:59.000Z

42

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

SciTech Connect

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

Kevin Whitty

2008-06-30T23:59:59.000Z

43

Black liquor gasification combined cycle with Co2 capture – Technical and economic analysis  

Science Journals Connector (OSTI)

Abstract The pulp and paper sector is intensive in the use of energy, and presents a high participation in the industrial context, specially based in the black liquor, a renewable source generated in the pulp process. Black liquor gasification is not still completely dominated; however, it has the potential of becoming an important alternative for the pulp and paper sector. In this article, the traditional steam cycle based on chemical recovery and biomass boilers associated to backpressure/extraction turbine is compared to black liquor gasification combined cycle schemes, associated to biomass boiler, considering the technical and economic attractiveness of capturing and sequestering CO2. Results show that despite its interesting exergetic efficiency, the adoption CO2 capture system for BLGCC did not prove to be attractive under the prescribed conditions without major incentive.

Elzimar Tadeu de Freitas Ferreira; José Antonio Perrella Balestieri

2014-01-01T23:59:59.000Z

44

Formation of fuel NO[sub x] during black-liquor combustion  

SciTech Connect

Fuel NOx and thermal NOx were measured in combustion gases from black liquors in two laboratory furnaces. Combustion at 950 C in air (8% O[sub 2]) produced NOx concentrations of 40-80ppm. Combustion at 950 C in synthetic air containing no nitrogen (21% 0[sub 2] in Ar) produced the same result, demonstrating that all of the NOx produced during combustion at 950 C was fuel NOx. Formation of fuel NOx increased moderately with increasing temperature in the range of 800-1,000 C, but temperature sensitivity of fuel NOx was much less than that of thermal NOx. The results imply that the major source of NOx in recovery furnace emissions is the fuel NOx in recovery furnace formed by conversion of liquor-bound nitrogen during combustion. This is consistent with thermal NOx theory, which postulates that black-liquor combustion temperatures are too low to generate significant amounts of thermal NOx.

Nichols, K.M. (Weyerhaeuser Paper Co., Tacoma, WA (United States)); Lien, S.J. (Inst. of Paper Science and Technology, Atlanta, GA (United States))

1993-03-01T23:59:59.000Z

45

A comprehensive program to develop correlations for the physical properties of Kraft black liquor. Interim report No. 2  

SciTech Connect

Experimental effort for the program to evaluate physical properties of kraft black liquors is now proceeding well. Experimental work includes pulping, liquor analysis, lignin purification and characterization, vapor-liquid equilibria, heat capacity, heats of solution and combustion, and viscosity measurements. Measurement of thermal conductivity has not yet begun. Collection of the data necessary for development of generalized correlations is proceeding, but will require about two more years. The digester is operating very well. It is now possible to operate the digester as a closed, rotating reactor or as a batch reactor with liquor circulation. When operated with liquor circulation, temperatures within the chip bed can be monitored during cooking. Cooking is reproducible, and cooks are being performed to produce liquors for experimental studies. The digester could be further modified to permit us to conduct rapid exchange batch pulping or to permit us to simulate continuous pulping. Liquors to be used in experimental studies are concentrated in our large scale evaporator or in our small scale evaporator. The large scale evaporator is used to concentrate liquors to about 50% solids for storage and for use in studies requiring high solids liquors. The small scale evaporator is used for preparing final samples to as high as 85% solids and for measuring vapor-liquid equilibria. Liquors are now routinely analyzed to determine all components, except higher molecular weight organic acids and extractives. Lignin determination by uv-visible means has been improved. Lignin purification from black liquor has been improved and lignin molecular weights are determined routinely. Work on lignin molecular weight distribution is still not satisfactory, but recent developments holds promise.

Fricke, A.L.

1990-12-01T23:59:59.000Z

46

Nuclear Regulatory Commission's Integrated Strategy for Spent...  

Office of Environmental Management (EM)

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management * 20+ years of...

47

Fast high-pressure freezing of protein crystals in their mother liquor  

Science Journals Connector (OSTI)

Protein crystals were vitrified using high-pressure freezing in their mother liquor at 210 MPa and 77 K without cryoprotectants or oil coating. The method was successfully applied to photosystem II, which is representative of a membrane protein with a large unit cell and weak crystal contacts.

Burkhardt, A.

2012-03-31T23:59:59.000Z

48

Final technical report. In-situ FT-IR monitoring of a black liquor recovery boiler  

SciTech Connect

This project developed and tested advanced Fourier transform infrared (FT-IR) instruments for process monitoring of black liquor recovery boilers. The state-of-the-art FT-IR instruments successfully operated in the harsh environment of a black liquor recovery boiler and provided a wealth of real-time process information. Concentrations of multiple gas species were simultaneously monitored in-situ across the combustion flow of the boiler and extractively at the stack. Sensitivity to changes of particulate fume and carryover levels in the process flow were also demonstrated. Boiler set-up and operation is a complex balance of conditions that influence the chemical and physical processes in the combustion flow. Operating parameters include black liquor flow rate, liquor temperature, nozzle pressure, primary air, secondary air, tertiary air, boiler excess oxygen and others. The in-process information provided by the FT-IR monitors can be used as a boiler control tool since species indicative of combustion efficiency (carbon monoxide, methane) and pollutant emissions (sulfur dioxide, hydrochloric acid and fume) were monitored in real-time and observed to fluctuate as operating conditions were varied. A high priority need of the U.S. industrial boiler market is improved measurement and control technology. The sensor technology demonstrated in this project is applicable to the need of industry.

James Markham; Joseph Cosgrove; David Marran; Jorge Neira; Chad Nelson; Peter Solomon

1999-05-31T23:59:59.000Z

49

Oil palm vegetation liquor: a new source of phenolic bioactives Ravigadevi Sambanthamurthi1  

E-Print Network (OSTI)

Oil palm vegetation liquor: a new source of phenolic bioactives Ravigadevi Sambanthamurthi1 *, Yew , Krishnan Subramaniam5 , Soon-Sen Leow1 , Kenneth C. Hayes6 and Mohd Basri Wahid1 1 Malaysian Palm Oil Board, 6, Persiaran Institusi, Bandar Baru Bangi, 43000 Kajang Selangor, Malaysia 2 Malaysian Palm Oil

Sinskey, Anthony J.

50

Black liquor combustion validated recovery boiler modeling, five-year report  

SciTech Connect

The objective of this project was to develop 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 project originated in October 1990 and was scheduled to run for four years. At that time, there was considerable emphasis on developing accurate predictions of the physical carryover of macroscopic particles of partially burnt black liquor and smelt droplets out of the furnace, since this was seen as the main cause of boiler plugging. This placed a major emphasis on gas flow patterns within the furnace and on the mass loss rates and swelling and shrinking rates of burning black liquor drops. As work proceeded on developing the recovery boiler furnace model, it became apparent that some recovery boilers encounter serious plugging problems even when physical carryover was minimal. After the original four-year period was completed, the project was extended to address this issue. The objective of the extended project was to improve the utility of the models by including the black liquor chemistry relevant to air emissions predictions and aerosol formation, and by developing the knowledge base and computational tools to relate furnace model outputs to fouling and plugging of the convective sections of the boilers. The work done to date includes CFD model development and validation, acquisition of information on black liquor combustion fundamentals and development of improved burning models, char bed model development, and model application and simplification.

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

1996-08-01T23:59:59.000Z

51

HFIR spent fuel management alternatives  

SciTech Connect

The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

1992-10-15T23:59:59.000Z

52

HFIR spent fuel management alternatives  

SciTech Connect

The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems` Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

1992-10-15T23:59:59.000Z

53

Interrelated Effects of Aeration and Mixed Liquor Fractions on Membrane Fouling for Submerged Membrane Bioreactor Processes in Wastewater Treatment  

Science Journals Connector (OSTI)

Interrelated Effects of Aeration and Mixed Liquor Fractions on Membrane Fouling for Submerged Membrane Bioreactor Processes in Wastewater Treatment ... using hollow fibers was applied to wastewater treated by the activated-sludge process. ...

Fengshen Fan; Hongde Zhou

2007-02-21T23:59:59.000Z

54

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

Office of Environmental Management (EM)

Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel More Documents &...

55

Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...  

Office of Environmental Management (EM)

Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives...

56

Spent-fuel-storage alternatives  

SciTech Connect

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)

Not Available

1980-01-01T23:59:59.000Z

57

Spent Fuel Disposal Trust Fund (Maine)  

Energy.gov (U.S. Department of Energy (DOE))

Any licensee operating a nuclear power plant in this State shall establish a segregated Spent Nuclear Fuel Disposal Trust Fund in accordance with this subchapter for the eventual disposal of spent...

58

Plutonium and Reprocessing of Spent Nuclear Fuel  

Science Journals Connector (OSTI)

...Repository for the Disposal of Spent Nuclear...Radioactive Waste at Yucca Mountain (YMP-0106...not committed funding to build...Repository for the Disposal of Spent Nuclear...Radioactive Waste at Yucca Mountain (YMP-0106, Yucca Mountain Project, North...

Frank N. von Hippel

2001-09-28T23:59:59.000Z

59

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

SciTech Connect

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.

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

1993-12-01T23:59:59.000Z

60

Computational Fluid Dynamics Simulations of Raw Gas Composition from a Black Liquor Gasifier—Comparison with Experiments  

Science Journals Connector (OSTI)

Computational Fluid Dynamics Simulations of Raw Gas Composition from a Black Liquor Gasifier—Comparison with Experiments ... The black liquor spray was represented by 1003 discrete particles having a fitted Rosin Rammler distribution of power 2 and a characteristic size of 200 ?m. ... Additional PFR calculations were performed using different inlet gas compositions with similar results (not presented here), i.e. the Jones and Lindstedt model showed a significantly higher reformation rate of methane than GRI-Mech at gasification conditions in the medium temperature range (1150 < T < 1500 K). ...

Per Carlsson; Kristiina Iisa; Rikard Gebart

2011-07-05T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Material Characterization and Analysis for Selection of Refractories Used in Black Liquor Gasification  

SciTech Connect

Black liquor gasification provides the pulp and paper industry with a technology which could potentially replace recovery boilers with equipment that could reduce emissions and, if used in a combined cycle system, increase the power production of the mill allowing it to be a net exporter of electrical power. In addition, rather than burning the syngas produced in a gasifier, this syngas could be used to produce higher value chemicals or fuels. However, problems with structural materials such as the refractory lining of the reactor vessel have caused unplanned shutdowns and resulted in component replacement much sooner than originally planned. Through examination of exposed materials, laboratory corrosion tests and cooperative efforts with refractory manufacturers, many refractory materials issues in high-temperature black liquor gasification have been addressed and optimized materials have been selected for this application. In this paper, an updated summary of the characterization and analysis techniques used for refractory screening and selection will be discussed along with characteristic results from these methods which have led to the selection of optimized materials for both the hot-face and back-up linings used in this application.

Hemrick, James Gordon [ORNL; Keiser, James R [ORNL; Meisner, Roberta A [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL)

2010-01-01T23:59:59.000Z

62

Spent Fuel Transportation Risk Assessment  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fuel Transportation Risk Assessment Fuel Transportation Risk Assessment (SFTRA) Draft NUREG-2125 Overview for National Transportation Stakeholders Forum John Cook Division of Spent Fuel Storage and Transportation 1 SFTRA Overview Contents * Project and review teams * Purpose and goals * Basic methodology * Improvements relative to previous studies * Draft NUREG structure and format * Routine shipment analysis and results * Accident condition analysis and results * Findings and conclusions * Schedule 2 SFTRA Research and Review Teams * Sandia National Laboratory Research Team [$1.8M; 9/06-9/12] - Doug Ammerman - principal investigator - Carlos Lopez - thermal - Ruth Weiner - RADTRAN * NRC's SFTRA Technical Review Team - Gordon Bjorkman - structural

63

Production of ammonium sulfate fertilizer from FGD waste liquors. First quarterly technical report, [January--March 1995  

SciTech Connect

Hydrolysis of Nitrogen-Sulfur Containing Compounds (N-SCC) derived from desulfurization liquors was carried out at high temperature and pressure with varying sulfuric acid concentration in order to determine the influence of temperature and acid concentration on time required for complete hydrolysis. An ammonia specific electrode (gas sensing) was used to monitor the concentration of ammonium ion in the hydrolyzed liquor. The results indicated a large shortening of time for complete hydrolysis of the Ca salt of N-SCC but varying acid concentration did influence the rate of hydrolysis. Since the physical-chemical analysis of the N-SCC obtained by re-liming the waste scrubbing liquor (containing the Fe-EDTA complex) was found to contain a high quantity of sodium, the N-SCC is believed to be a double salt of calcium and sodium. The final product, (NH{sub 4}){sub 2}SO{sub 4}, was obtained in experimentation using an evaporative crystallizer. The clean hydrolyzed liquor was neutralized with a strong NH{sub 2} solution before the final crystallization of (NH{sub 4}){sub 2}SO{sub 4}.

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

1995-08-01T23:59:59.000Z

64

Cracking and Corrosion of Composite Tubes in Black Liquor Recovery Boiler Primary Air Ports  

SciTech Connect

Black liquor recovery boilers are an essential part of kraft mills. Their design and operating procedures have changed over time with the goal of providing improved boiler performance. These performance improvements are frequently associated with an increase in heat flux and/or operating temperature with a subsequent increase in the demand on structural materials associated with operation at higher temperatures and/or in more corrosive environments. Improvements in structural materials have therefore been required. In most cases the alternate materials have provided acceptable solutions. However, in some cases the alternate materials have solved the original problem but introduced new issues. This report addresses the performance of materials in the tubes forming primary air port openings and, particularly, the problems associated with use of stainless steel clad carbon steel tubes and the solutions that have been identified.

Keiser, James R.; Singbeil, Douglas L.; Sarma, Gorti B.; Kish, Joseph R.; Yuan, Jerry; Frederick, Laurie A.; Choudhury, Kimberly A.; Gorog, J. Peter; Jetté, Francois R.; Hubbard, Camden R.; Swindeman, Robert W.; Singh, Prett M.; Maziasz, Phillip J.

2006-10-01T23:59:59.000Z

65

Spent Nuclear Fuel (SNF) Project Execution Plan  

SciTech Connect

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.

LEROY, P.G.

2000-11-03T23:59:59.000Z

66

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

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

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...

67

TEPP - Spent Nuclear Fuel | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

- Spent Nuclear Fuel - Spent Nuclear Fuel TEPP - Spent Nuclear Fuel 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 a series of five scenarios developed by the Department of Energy Transportation Emergency Preparedness Program. Responding agencies may include several or more of the following: local municipal and county fire, police, sheriff, and Emergency Medical Services (EMS) personnel; state, local, and federal emergency response teams; emergency response contractors;and other emergency response resources that could potentially be provided by the carrier and the originating facility (shipper). Spent Nuclear Fuel.docx More Documents & Publications

68

Recovery of manganese oxides from spent alkaline and zinc–carbon batteries. An application as catalysts for VOCs elimination  

SciTech Connect

Highlights: • Manganese oxides were synthesized using spent batteries as raw materials. • Spent alkaline and zinc–carbon 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 zinc–carbon 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 zinc–carbon batteries as raw materials, have an interesting performance as catalysts for elimination of VOCs.

Gallegos, María V., E-mail: plapimu@yahoo.com.ar [Pla.Pi.Mu-Planta Piloto Multipropósito, (CICPBA-UNLP) Cno. Centenario y 505, M.B. Gonnet, Buenos Aires (Argentina); Falco, Lorena R., E-mail: mlfalco@quimica.unlp.edu.ar [Pla.Pi.Mu-Planta Piloto Multipropósito, (CICPBA-UNLP) Cno. Centenario y 505, M.B. Gonnet, Buenos Aires (Argentina); Peluso, Miguel A., E-mail: apelu@quimica.unlp.edu.ar [Centro de Investigación y Desarrollo en Ciencias Aplicadas, “Dr. J. Ronco” CINDECA (CONICET CCT La Plata), 47 N°257, La Plata, Buenos Aires (Argentina); Sambeth, Jorge E., E-mail: sambeth@quimica.unlp.edu.ar [Centro de Investigación y Desarrollo en Ciencias Aplicadas, “Dr. J. Ronco” CINDECA (CONICET CCT La Plata), 47 N°257, La Plata, Buenos Aires (Argentina); Thomas, Horacio J. [Pla.Pi.Mu-Planta Piloto Multipropósito, (CICPBA-UNLP) Cno. Centenario y 505, M.B. Gonnet, Buenos Aires (Argentina)

2013-06-15T23:59:59.000Z

69

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

Office of Environmental Management (EM)

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...

70

2008 DOE Spent Nuclear Fuel and High Level Waste Inventory  

NLE Websites -- All DOE Office Websites (Extended Search)

Management >> National Spent Nuclear Fuel INL Logo Search 2008 DOE Spent Nuclear Fuel and High Level Waste Inventory Content Goes Here Skip Navigation Links Home Newsroom About INL...

71

Safeguards for spent fuels: Verification problems  

SciTech Connect

The accumulation of large quantities of spent nuclear fuels world-wide is a serious problem for international safeguards. A number of International Atomic Energy Agency (IAEA) member states, including the US, consider spent fuel to be a material form for which safeguards cannot be terminated, even after permanent disposal in a geologic repository. Because safeguards requirements for spent fuels are different from those of conventional bulk-handling and item-accounting facilities, there is room for innovation to design a unique safeguards regime for spent fuels that satisfies the goals of the nuclear nonproliferation treaty at a reasonable cost to both the facility and the IAEA. Various strategies being pursued for long-term management of spent fuels are examined with a realistic example to illustrate the problems of verifying safeguards under the present regime. Verification of a safeguards regime for spent fuels requires a mix of standard safeguards approaches, such as quantitative verification and use of seals, with other measures that are unique to spent fuels. 17 refs.

Pillay, K.K.S.; Picard, R.R.

1991-01-01T23:59:59.000Z

72

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

E-Print Network (OSTI)

hydraulic cement from spent oil shale," Vol. 10, No. 4, p.J. W. , "Colorado's primary oil shale resource for verticalSimulated effects of oil-shale development on the hydrology

Mehta, P.K.

2013-01-01T23:59:59.000Z

73

Spent fuel storage requirements 1993--2040  

SciTech Connect

Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges.

Not Available

1994-09-01T23:59:59.000Z

74

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

SciTech Connect

Low cost material is needed for grouting abandoned retorts. Experimental work has shown that a hydraulic cement can be produced from Lurgi spent shale by mixing it in a 1:1 weight ratio with limestone and heating one hour at 1000°C. With 5% added gypsum, strengths up to 25.8 MPa are obtained. This cement could make an economical addition up to about 10% to spent shale grout mixes, or be used in ordinary cement applications.

Mehta, P.K.; Persoff, P.; Fox, J.P.

1980-06-01T23:59:59.000Z

75

Pyrochemical processing of DOE spent nuclear fuel  

SciTech Connect

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.

Laidler, J.J.

1995-02-01T23:59:59.000Z

76

Gasification behavior of carbon residue in bed solids of black liquor gasifier  

SciTech Connect

Steam gasification of carbon residue in bed solids of a low-temperature black liquor gasifier was studied using a thermogravimetric system at 3 bar. Complete gasification of the carbon residue, which remained unreactive at 600 C, was achieved in about 10 min as the temperature increased to 800 C. The rate of gasification and its temperature dependence were evaluated from the non-isothermal experiment results. Effects of particle size and adding H{sub 2} and CO to the gasification agent were also studied. The rate of steam gasification could be taken as zero order in carbon until 80% of carbon was gasified, and for the rest of the gasification process the rate appeared to be first order in carbon. The maximum rate of carbon conversion was around 0.003/s and the activation energy was estimated to be in the range of 230-300 kJ/mol. The particle size did not show significant effect on the rate of gasification. Hydrogen and carbon monoxide appeared to retard the onset of the gasification process. (author)

Preto, Fernando; Zhang, Xiaojie (Frank); Wang, Jinsheng [CANMET Energy Technology Centre, Natural Resources (Canada)

2008-07-15T23:59:59.000Z

77

Spent nuclear fuel reprocessing modeling  

SciTech Connect

The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V. [Bochvar Institute, 5 Rogova str., Moscow 123098 (Russian Federation); Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I. [Russian Federal Nuclear Center - VNIITF E.I. Zababakhin, p.o.box 245, Snezhinsk, 456770 (Russian Federation)

2013-07-01T23:59:59.000Z

78

Spent Nuclear Fuel Alternative Technology Decision Analysis  

SciTech Connect

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.

Shedrow, C.B.

1999-11-29T23:59:59.000Z

79

Spent Fuel Background Report Volume I  

SciTech Connect

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.

Abbott, D.

1994-03-01T23:59:59.000Z

80

The intoxicating brew of black liquor and son of black liquor: Deciphering the tax planning, research, policy and financial statement implications of tax credits via financial reporting income tax disclosures  

Science Journals Connector (OSTI)

In 2007, a change in the law regarding the alternative fuel mixture credit opened the door for paper mills to qualify a byproduct of paper manufacturing, black liquor, as a fuel eligible for the credit. The credit is a refundable credit of $0.50 per gallon. Paper mills can produce hundreds of millions of gallons of black liquor per year and qualified for the credit in 2009. In addition, in 2010 the IRS determined that these firms qualified for the cellulosic biofuels producer credit. Paper mill companies could amend their 2009 tax returns and swap their alternative fuel mixture credits for cellulosic biofuels producer credits worth $1.01 per gallon. The catch was that the alternative fuels mixture credit was refundable; the cellulosic biofuels producer credit was nonrefundable. This paper provides a series of mini cases exploring the tax planning, tax research, tax policy and financial statement implications of Packaging Corporation of America’s use of the alternative fuel mixture tax credit and subsequent amendment of its 2009 tax return in 2010 to swap its alternative fuel mixture tax credits for cellulosic biofuel producer credits. These cases may be directed for use in either upper-division undergraduate courses or graduate courses.

Beth B. Kern

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Optimization study on sample pretreatment of spent fuel storage rack  

Science Journals Connector (OSTI)

In order to evaluate radionuclide inventories as an essential item for the permanent disposal of spent fuel storage racks, chemical conditions for a sample pretreatment of a spent fuel storage rack were studied. ...

Hong-Joo Ahn; Myung-Ho Lee; Se-Chul Sohn…

2010-08-01T23:59:59.000Z

82

INVESTIGATIONS ON HYDRAULIC CEMENTS FROM SPENT OIL SHALE  

E-Print Network (OSTI)

CEMENTS FROM SPENT OIL SHALE P.K. Mehta and P. Persoff AprilCement Manufacture from Oil Shale, U.S. Patent 2,904,445,CEMENTS FROM SPENT OIL SHALE P, K, Mehta Civil Engineering

Mehta, P.K.

2012-01-01T23:59:59.000Z

83

EIS-0015: U.S. Spent Fuel Policy  

Energy.gov (U.S. Department of Energy (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.

84

Spent Sealed Sources Management in Switzerland - 12011  

SciTech Connect

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)

Beer, H.F. [Paul Scherrer Institute, CH-5232 Villigen (Switzerland)

2012-07-01T23:59:59.000Z

85

Fundamental study of black liquor gasification kinetics using a pressurized entrained-flow reactor (PEFR). Quarterly progress report for the period July 1999 to September 1999  

SciTech Connect

The goal of the program is to identify the optimal operating window for black liquor gasification. The goals during this year are to prepare the PEFR for operation, conduct a series of preliminary screening tests to bracket BLG operating conditions, and develop a process model that can guide identification of the optimal operating window.

NONE

1999-10-29T23:59:59.000Z

86

Mineral phases of green liquor dregs, slaker grits, lime mud and wood ash of a Kraft pulp and paper mill  

Science Journals Connector (OSTI)

Four residues generated in a Kraft, pulp and paper plant, were characterized by X-ray fluorescence spectroscopy (XFA), powder X-ray diffraction (XRD), thermogravimmetric analysis (TG) and Fourier transform infrared spectroscopy (FTIR). A quantitative phase composition model, that accounts for the observed data and for the physico-chemical conditions of formation, was postulated for each material. Emphasis was given on the identification of the mineral components of each material. The green liquor dregs and the lime mud contain Calcite and Gipsite. The slaker grits contains Calcite, Portlandite, Pirssonite, Larnite and Brucite. The Calcite phase, present in the dregs and in the lime mud, has small amounts of magnesium replacing calcium. The wood ash contains Quartz as the major crystalline mineral phase.

Fernanda Machado Martins; Joaniel Munhoz Martins; Luiz Carlos Ferracin; Carlos Jorge da Cunha

2007-01-01T23:59:59.000Z

87

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

SciTech Connect

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.

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

1994-12-31T23:59:59.000Z

88

Microsoft Word - spent nuclear fuel report.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management of Spent Nuclear Fuel Management of Spent Nuclear Fuel at the Savannah River Site DOE/IG-0727 May 2006 REPORT ON MANAGEMENT OF SPENT NUCLEAR FUEL AT THE SAVANNAH RIVER SITE TABLE OF CONTENTS Spent Nuclear Fuel Management Details of Finding 1 Recommendations 2 Comments 3 Appendices 1. Objective, Scope, and Methodology 4 2. Prior Audit Reports 5 3. Management Comments 6 SPENT NUCLEAR FUEL MANGEMENT Page 1 Details of Finding H-Canyon The Department of Energy's (Department) spent nuclear fuel Operations program at the Savannah River Site (Site) will likely require Extended H-Canyon to be maintained at least two years beyond defined operational needs. The Department committed to maintain H-Canyon operational readiness to provide a disposal path for

89

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

79: Spent Nuclear Fuel Management, Aiken, South Carolina 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 OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 5, 2013 EIS-0279: Amended Record of Decision Spent Nuclear Fuel Management at the Savannah River Site April 1, 2013 EIS-0279-SA-01: Supplement Analysis Savannah River Site Spent Nuclear Fuel Management (DOE/EIS-0279-SA-01 and

90

Naval Spent Fuel Rail Shipment Accident Exercise Objectives  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NAVAL SPENT FUEL RAIL SHIPMENT 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 y p simulated accident * Gain understanding of how the communications links that would be activated in an accident involving a Naval spent fuel shipment would work 1 NTSF May 11 ACCIDENT EXERCISE TYPICAL TIMELINE * Conceptual/Organizational Meeting - April 6 E R T i d it t t d TYPICAL TIMELINE

91

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0203: Spent Nuclear Fuel Management and Idaho National EIS-0203: 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 projected quantities of spent nuclear fuel until the year 2035. This amount of time may be required to make and implement a decision on the ultimate disposition of spent nuclear fuel. DOE's spent nuclear fuel responsibilities include fuel generated by DOE production, research, and development reactors; naval reactors; university and foreign research reactors; domestic non-DOE reactors such as those at the National Institute

92

DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER  

SciTech Connect

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.

F. Habashi

1998-06-26T23:59:59.000Z

93

Surrogate Spent Nuclear Fuel Vibration Integrity Investigation  

SciTech Connect

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.

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

2014-01-01T23:59:59.000Z

94

Pyrochemical Treatment of Spent Nuclear Fuel  

SciTech Connect

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 Energy’s 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.

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

2005-10-01T23:59:59.000Z

95

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

96

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

National Nuclear Security Administration (NNSA)

Spent Fuel Storage Facility Washington, DC The Reagan Administration announces a nuclear energy policy that anticipates the establishment of a facility for the storage of...

97

Transportation capabilities study of DOE-owned spent nuclear fuel  

SciTech Connect

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01T23:59:59.000Z

98

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

SciTech Connect

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)

Jardine, Leslie J. [L. J. Jardine Services, Consultant, Dublin CA, 94568 (United States)

2012-07-01T23:59:59.000Z

99

Radiological characterization of spent control rod assemblies  

SciTech Connect

This document represents the final report of an ongoing study to provide radiological characterizations, classifications, and assessments in support of the decommissioning of nuclear power stations. This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determined by three separate methodologies, including (1) direct assay techniques, (2) calculational techniques, and (3) by sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assembly (BPRA), {sup 60}Co and {sup 63}Ni, present in the stainless steel cladding, were the most abundant neutron activation products. The most abundant radionuclide in the PWR rod cluster control assembly (RCCA) was {sup 108m}Ag (130 yr halflife) produced in the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contributor to the gamma dose rate for many hundreds of years. The results of the direct assay methods agree very well ({+-}10%) with the sampling/radiochemical measurements. The results of the calculational methods agreed fairly well with the empirical measurements for the BPRA, but often varied by a factor of 5 to 10 for the CRB and the RCCA assemblies. If concentration averaging and encapsulation, as allowed by 10CFR61.55, is performed, then each of the entire control assemblies would be classified as Class C low-level radioactive waste.

Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L. [Pacific Northwest Lab., Richland, WA (United States)

1995-10-01T23:59:59.000Z

100

Breeder Spent Fuel Handling Program multipurpose cask design basis document  

SciTech Connect

The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

Duckett, A.J.; Sorenson, K.B.

1985-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Testing of the CANDU Spent Fuel Storage Basket Package  

SciTech Connect

The paper described the results of testing for a CANDU Spent Fuel Storage Basket Package Prototype intended to be used for transport and storage of the CANDU spent fuel bundles within NPP CANDU Cernavoda, Romania. The results obtained proved that the objectives of those tests were achieved

Vieru, G.

2002-02-28T23:59:59.000Z

102

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

SciTech Connect

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.

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

1998-08-01T23:59:59.000Z

103

Characterization plan for Hanford spent nuclear fuel  

SciTech Connect

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.

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

1994-12-01T23:59:59.000Z

104

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

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

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...

105

Loss of spent fuel pool cooling PRA: Model and results  

SciTech Connect

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.

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

1996-09-01T23:59:59.000Z

106

DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS  

SciTech Connect

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.

Rudisill, T; John Mickalonis, J

2006-09-27T23:59:59.000Z

107

INVESTIGATIONS ON HYDRAULIC CEMENTS FROM SPENT OIL SHALE  

SciTech Connect

A process for making hydraulic cements from spent oil shale is described in this paper. Inexpensive cement is needed to grout abandoned in-situ retorts of spent shale for subsidence control, mitigation of leaching, and strengthening the retorted mass in order to recover oil from adjacent pillars of raw shale. A hydraulic cement was produced by heating a 1:1 mixture of Lurgi spent shale and CaCO{sub 3} at 1000 C for one hour. This cement would be less expensive than ordinary portland cement and is expected to fulfill the above requirements.

Mehta, P.K.; Persoff, P.

1980-04-01T23:59:59.000Z

108

INVESTIGATIONS ON HYDRAULIC CEMENTS FROM SPENT OIL SHALE  

E-Print Network (OSTI)

20 to 40% of the oil shale, and explosively rubblizing andCEMENTS FROM SPENT OIL SHALE P.K. Mehta and P. Persoff AprilCement Manufacture from Oil Shale, U.S. Patent 2,904,445,

Mehta, P.K.

2012-01-01T23:59:59.000Z

109

Risk and Responsibility Sharing in Nuclear Spent Fuel Management  

E-Print Network (OSTI)

With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

De Roo, Guillaume

110

President Reagan Calls for a National Spent Fuel Storage Facility |  

National Nuclear Security Administration (NNSA)

Reagan Calls for a National Spent Fuel Storage Facility | Reagan Calls for a National Spent Fuel Storage Facility | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > President Reagan Calls for a National Spent ... President Reagan Calls for a National Spent Fuel Storage Facility October 08, 1981

111

Extraction of uranium from spent fuels using liquefied gases  

SciTech Connect

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)

Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan)

2007-07-01T23:59:59.000Z

112

Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards  

E-Print Network (OSTI)

on Spent Fuel for Nuclear Safeguards Brian J. Quiter, ?resonances on nuclear safeguards measurements will be

Quiter, Brian

2013-01-01T23:59:59.000Z

113

The united kingdom's changing requirements for spent fuel storage  

SciTech Connect

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.

Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N. [National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, Cheshire WA3 6AE (United Kingdom)

2013-07-01T23:59:59.000Z

114

Conditioning of spent nuclear fuel for permanent disposal  

SciTech Connect

A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, 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 US 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 that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

Laidler, J.J. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

115

Conditioning of spent nuclear fuel for permanent disposal  

SciTech Connect

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 preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

Laidler, J.J.

1994-10-01T23:59:59.000Z

116

CASMO-2 spent-fuel-rack criticality analysis  

SciTech Connect

In recent years, utilities have needed to increase their spent-fuel storage capacity. Both Maine Yankee pressurized water reactor (PWR) and Vermont Yankee boiling water reactor (BWR) have increased their spent-fuel rack capacity by decreasing the canister center-to-center spacing while adding fixed poison. Licensing criticality analysis of such changes in spent-fuel rack design have been performed at Yankee Atomic Electric Co. (YAEC) using NITAWL-KENO-IV and the 123-group XSDRN library. However, KENO/Monte Carlo analysis has inherent drawbacks when applied to spent-fuel rack design and modification. These include statistical uncertainty and long computer time. In contrast, the transport theory code, CASMO-2, provides deterministic and fast criticality analysis. Also, since collapsed and transport-corrected cross sections are generated, PDQ can be used to analyze large array problems which are prohibitively expensive using KENO. In this work, the authors apply the CASMO-PDQ methodology to the Maine Yankee and Vermont Yankee high-density spent-fuel rack designs, and compare the final results against KENO.

Napolitano, D.G.; Heinrichs, D.P.; Gorski, J.P.

1986-01-01T23:59:59.000Z

117

Pyroprocessing oxide spent nuclear fuels for efficient disposal  

SciTech Connect

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment.

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

118

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NRC's 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 Growing Spent Fuel Inventory Cumulative Used Nuclear Fuel Scenarios 50,000 100,000 150,000 200,000 250,000 Metric Tons 3 - 50,000 2010 2015 2020 2025 2030 2035 2040 2045 2050 Year Reference: Crozat, March 2010 Integrated Strategy * In response to the evolving national debate on spent fuel management strategy, NRC initiated a number of actions:

119

Spent fuel storage system for LMFBR fuel experiments  

SciTech Connect

Fuel that had been irradiated in the Argonne National Laboratory Experimental Breeder Reactor II (EBR-II) at Idaho Falls, Idaho, and examined at the Hanford Engineering Development Laboratory at Richland, Washington, was placed in long term retrievable storage utilizing a system designed at Hanford. The Spent Fuel Storage Cask system was designed for transport and storage of a large quantity of spent fuel at the Hanford 200 Area transuranic (TRU) asphalt storage pad. The entire system is designed for long term retrievable storage to allow future reprocessing of the fuel. The system was designed to meet the criticality, shielding, and thermal requirements for a maximum fuel load of four kilograms fissile. The Spent Fuel Storage Cask was built to transport and store the fuel from EBR-II on the TRU asphalt storage pad.

Seay, J.M.; Gruber, W.J.

1983-01-01T23:59:59.000Z

120

Information handbook on independent spent fuel storage installations  

SciTech Connect

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.

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

1996-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy (DOE) radioactive wastes were compiled through December 31, 1983, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 37 years and characteristics of these materials are also presented, consistent with the latest DOE/Energy Information Administration (EIA) or projection of US commercial nuclear power growth and expected defense-related and private industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. 48 figures, 107 tables.

Not Available

1984-09-01T23:59:59.000Z

122

SEU43 fuel bundle shielding analysis during spent fuel transport  

SciTech Connect

The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

Margeanu, C. A.; Ilie, P.; Olteanu, G. [Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)

2006-07-01T23:59:59.000Z

123

Dry storage of spent nuclear fuel in UAE – Economic aspect  

Science Journals Connector (OSTI)

Abstract Cost analysis of dry storage of spent nuclear fuel (SNF) discharged from Barakah nuclear power plants in the UAE was performed using three variables: average fuel discharge rate (FD), discount rate (d), and cooling time in a spent fuel pool (Tcool). The costs of dry storage as an interim spent fuel storage option in the UAE were estimated and compared between the following two scenarios: Scenario 1 is ‘accelerated transfer of spent fuel to dry storage’ that SNF will be transferred to dry storage facilities as soon as spent fuel has been sufficiently cooled down in a pool for the dry storage; Scenario 2 is defined as ‘maximum use of spent fuel pool’ that SNF will be stored in a pool as long as possible till the amount of stored SNF in the pool reaches the capacity of the pools and, then, to be moved to dry storage. A sensitivity analysis on the costs was performed and multiple regression analysis was applied to the resulting net present values (NPVs) for Scenarios 1 and 2 and ?NPV that is difference in the net present values between the two scenarios. The results showed that \\{NPVs\\} and ?NPV could be approximately expressed by single equations with the three variables. Among the three variables, the discount rate had the largest effect on the \\{NPVs\\} of the dry storage costs. However, ?NPV was turned out to be equally sensitive to the discount rate and cooling period. Over the ranges of the variables, the additional cost for accelerated fuel transfer (Scenario 1) ranged from 86.4 to 212.9 million $. Calculated using the maximum difference (212.9 M$) between the two scenarios, the accelerated fuel transfer to dry storage could incur the additional electricity rate 8.0 × 10?5 USD/kWh, which is not considered to be significant, compared to the overall electricity generation cost.

Sara Al Saadi; Yongsun Yi

2015-01-01T23:59:59.000Z

124

Spent fuel dissolution studies FY 1991 to 1994  

SciTech Connect

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.

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

1995-12-01T23:59:59.000Z

125

Method for the regeneration of spent molten zinc chloride  

DOE Patents (OSTI)

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.

Zielke, Clyde W. (McMurray, PA); Rosenhoover, William A. (Pittsburgh, PA)

1981-01-01T23:59:59.000Z

126

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect

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).

Hanson, B.D.

1998-07-01T23:59:59.000Z

127

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - III: Spent DUPIC Fuel Disposal Cost  

SciTech Connect

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Ko, Won Il; Choi, Hangbok; Roh, Gyuhong; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

128

MANAGEMENT OF SPENT NUCLEAR FUEL IN FINLAND: POLICY, PAST AND PRESENT PRACTICES, PLANS FOR THE FUTURE  

Science Journals Connector (OSTI)

In Finland, about 1700 tU of spent nuclear fuel has arisen from the operation of the four nuclear power units which were commissioned in late ... 1980’s. Initially the spent fuel management policy was based on se...

E. RUOKOLA

2006-01-01T23:59:59.000Z

129

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

Energy.gov (U.S. Department of Energy (DOE)) 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...

130

E-Print Network 3.0 - aluminum-clad spent nuclear Sample Search...  

NLE Websites -- All DOE Office Websites (Extended Search)

clad spent nuclear Search Powered by Explorit Topic List Advanced Search Sample search results for: aluminum-clad spent nuclear Page: << < 1 2 3 4 5 > >> 1 Hanford K-Basin Sludge...

131

E-Print Network 3.0 - aluminum-based spent nuclear Sample Search...  

NLE Websites -- All DOE Office Websites (Extended Search)

spent nuclear Search Powered by Explorit Topic List Advanced Search Sample search results for: aluminum-based spent nuclear Page: << < 1 2 3 4 5 > >> 1 Nuclear engineers design,...

132

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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...

133

An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly  

E-Print Network (OSTI)

Nuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of ...

Lovett, Phyllis Maria

1991-01-01T23:59:59.000Z

134

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2: Urgent-Relief Acceptance of Foreign Research Reactor Spent 2: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel SUMMARY 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 the ocean in spent fuel transportation casks from the country of origin to one or more United States eastern seaboard ports. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 22, 1994 EA-0912: Finding of No Significant Impact Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel April 22, 1994 EA-0912: Final Environmental Assessment Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

135

The Impacts of Dry-Storage Canister Designs on Spent Nuclear...  

Office of Environmental Management (EM)

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...

136

Methodology for determining criteria for storing spent fuel in air  

SciTech Connect

Dry storage in an air atmosphere is a method being considered for spent light water reactor (LWR) fuel as an alternative to storage in an inert gas environment. However, methods to predict fuel integrity based on oxidation behavior of the fuel first must be evaluated. The linear cumulative damage method has been proposed as a technique for defining storage criteria. Analysis of limited nonconstant temperature data on nonirradiated fuel samples indicates that this approach yields conservative results for a strictly decreasing-temperature history. On the other hand, the description of damage accumulation in terms of remaining life concepts provides a more general framework for making predictions of failure. Accordingly, a methodology for adapting remaining life concepts to UO/sub 2/ oxidation has been developed at Pacific Northwest Laboratory. Both the linear cumulative damage and the remaining life methods were used to predict oxidation results for spent fuel in which the temperature was decreased with time to simulate the temperature history in a dry storage cask. The numerical input to the methods was based on oxidation data generated with nonirradiated UO/sub 2/ pellets. The calculated maximum allowable storage temperatures are strongly dependent on the temperature-time profile and emphasize the conservatism inherent in the linear cumulative damage model. Additional nonconstant temperature data for spent fuel are needed to both validate the proposed methods and to predict temperatures applicable to actual spent fuel storage.

Reid, C.R.; Gilbert, E.R.

1986-11-01T23:59:59.000Z

137

Zircaloy cladding: Tough Containment for Spent Fuel Storage  

SciTech Connect

The present work leads us to strongly believe that if the cladding temperature is held below 425/sup 0/C then the cladding provides a significant barrier against radionuclide migration for over 1000 years. Unless some alternative benefit is identified, we feel that steps should be taken to preserve the integrity of the Zircaloy cladding - Tough Containment for Spent Fuel Storage.

Einziger, R.E.; Bosi, D.M.; Miller, A.K.

1981-01-01T23:59:59.000Z

138

Ceramic-composite waste forms from the electrometallurgical treatment of spent nuclear fuel  

Science Journals Connector (OSTI)

Argonne National Laboratory is developing a method to treat spent nuclear fuel in a molten-salt electrorefiner. Glass...

C. Pereira; M. Hash; M. Lewis; M. Richmann

1997-07-01T23:59:59.000Z

139

Corn steep liquor and fermented ammoniated condensed whey as protein sources for lactating cows and yearling heifers grazing winter native range  

SciTech Connect

Corn steep liquor (CSL) and fermented ammoniated condensed whey (FACW) were compared to cottonseed meal (CSM) as protein sources for wintering 61 lactating first-calf Hereford heifers and 32 yearling Hereford heifers on native range. Cattle were allotted by weight and individually fed 6 days per week for 12 weeks one of four protein treatments: negative control (NC), positive control (PC), CSL and FACW to provide .7, 1.5, .15 and 1.5 lb crude protein (CP) per day, respectively, to the lacating heifers and .2, .4, .4 and .4lb cP per day, respectively, to the yearling heifers. CMS was supplied in the CSL and FACW treatments at the same level as in the negative control. Lactating heifers fed the NC lost more (P less than .005) weight and body condition (120 lb and 1.6 units) than those fed the PC (45.8 lb and .9 units). Weight and condition losses were similar (P more than .05) for lactating heifers fed PC, CSL and FACW. Yearling heifers fed the NC lost more (P less than .005) weight than those fed the PC (49.4 vs 10.6 lb). Yearling heifers fed CSL and FACW gained more (P less than .005) weight than those fed the PC (17.6 and 9.3 vs - 10.6 lb). Feeding CSL resulted in signficantly lower rumen pH, lower ruminal acetate and higher ruminal butyrate, isovalerate and caproate levels than did feeding either control. Supplementing with FACW produced significantly lower rumen pH, higher rumen ammonia and soluble carbohydrate levels, lower ruminal acetate, and higher ruminal propionate and butyrate concentrations than did either control supplement. Corn steep liquor and FDCW appear to be effective protein sources for cows and heifers grazing winter native range.

Wagner, J.J.; Lusby, K.S.; Horn, G.W.; Dvorak, M.J.

1982-06-01T23:59:59.000Z

140

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

306: Treatment and Management of Sodium-Bonded Spent Nuclear 306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel Summary This EIS evaluates the potential environmental impacts of the proposed electrometallurgical treatment of DOE-owned sodium bonded spent nuclear fuel in the Fuel Conditioning Facility at Argonne National Laboratory-West (ANL-W). Public Comment Opportunities None available at this time. Documents Available for Download September 19, 2000 EIS-0306: Record of Decision Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 2000 EIS-0306: Final Environmental Impact Statement Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 1999 EIS-0306: Draft Environmental Impact Statement Treatment of Sodium-Bonded Spent Nuclear Fuel

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141

Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel  

SciTech Connect

Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O{sub 2}{sup 2+} mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin ({approx}20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U{sup 4+} environment. Available data for the standard reduction potentials for NpO{sup 2+}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote/enhance nucleation of NpO{sub 2} and Np{sub 2}O{sub 5}. Alternatively, Np may be incorporated into uranyl (UO{sub 2}{sup 2+}) alteration phases [2]. In some cases, less-soluble elements such as plutonium will be enriched near the surface of the corroding fuel [3]. We have used focused synchrotron x-rays from the MRCAT beam line at the Advanced Photon Source (APS) at Argonne National Lab to examine a specimen of spent nuclear fuel that had been subject to 10 years of corrosion testing in an environment of humid air and dripping groundwater at 90 C [4]. We find evidence of a region, approximately 20 microns in thickness, enriched in plutonium and neptunium at the corrosion front that exists between the uranyl silicate alteration mineral rind and the unaltered uranium oxide fuel (Figures 1 and 2). The uranyl silicate is itself found to be depleted in these transuranic elements relative to their abundance relative to uranium in the parent fuel. This suggests a low mobility of these components owing to a resistance to oxidize further in the presence of a UO{sub 2}{sup 2+}/U{sup 4+} couple [5].

J.A> Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

2006-06-20T23:59:59.000Z

142

Removal of arsenic compounds from spent catecholated polymer  

DOE Patents (OSTI)

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.

Fish, Richard H. (Berkeley, CA)

1985-01-01T23:59:59.000Z

143

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-06-06T23:59:59.000Z

144

Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography  

E-Print Network (OSTI)

This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

Jonkmans, G; Jewett, C; Thompson, M

2012-01-01T23:59:59.000Z

145

Gamma Ray Mirrors for Direct Measurement of Spent Nuclear Fuel  

SciTech Connect

Direct measurement of the amount of Pu and U in spent nuclear fuel represents a challenge for the safeguards community. Ideally, the characteristic gamma-ray emission lines from different isotopes provide an observable suitable for this task. However, these lines are generally lost in the fierce flux of radiation emitted by the fuel. The rates are so high that detector dead times limit measurements to only very small solid angles of the fuel. Only through the use of carefully designed view ports and long dwell times are such measurements possible. Recent advances in multilayer grazing-incidence gamma-ray optics provide one possible means of overcoming this difficulty. With a proper optical and coating design, such optics can serve as a notch filter, passing only narrow regions of the overall spectrum to a fully shielded detector that does not view the spent fuel directly. We report on the design of a mirror system and a number of experimental measurements.

Pivovaroff, Dr. Michael J. [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Ziock, Klaus-Peter [ORNL] [ORNL; Harrison, Mark J [ORNL] [ORNL; Soufli, Regina [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL)

2014-01-01T23:59:59.000Z

146

Prototype spent-fuel canister design, analysis, and test  

SciTech Connect

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included.

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01T23:59:59.000Z

147

Method For Processing Spent (Trn,Zr)N Fuel  

DOE Patents (OSTI)

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.

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

2004-07-27T23:59:59.000Z

148

Adsorption of phenol from aqueous systems onto spent oil shale  

SciTech Connect

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.

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

1996-03-01T23:59:59.000Z

149

Machine Vision Tests for Spent Fuel Scrap Characteristics  

SciTech Connect

The purpose of this work is to perform a feasibility test of a Machine Vision system for potential use at the Hanford K basins during spent nuclear fuel (SNF) operations. This report documents the testing performed to establish functionality of the system including quantitative assessment of results. Fauske and Associates, Inc., which has been intimately involved in development of the SNF safety basis, has teamed with Agris-Schoen Vision Systems, experts in robotics, tele-robotics, and Machine Vision, for this work.

BERGER, W.W.

2000-04-27T23:59:59.000Z

150

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

SciTech Connect

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.

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

151

Standard guide for drying behavior of spent nuclear fuel  

E-Print Network (OSTI)

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

152

Spent-fuel pool thermal hydraulics: The evaporation question  

SciTech Connect

Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40{degrees}C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas.

Yilmaz, T.P. [Belcan Services, Lombard, IL (United States); Lai, J.C. [Public Service Electric & Gas Co., Berwick, PA (United States)

1996-12-31T23:59:59.000Z

153

An approach to determine a defensible spent fuel ratio.  

SciTech Connect

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.

Durbin, Samuel G.; Lindgren, Eric Richard

2014-03-01T23:59:59.000Z

154

EIS-0245: Management of Spent Fuel from the K Basins at the Hanford Site -  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

245: Management of Spent Fuel from the K Basins at the Hanford 245: Management of Spent Fuel from the K Basins at the Hanford Site - Supplement Analysis, Richland, Washington EIS-0245: Management of Spent Fuel from the K Basins at the Hanford Site - Supplement Analysis, Richland, Washington Overview Overview to be provided. Public Comment Opportunities No public comment opportunities available at this time. Documents Available for Download August 15, 2011 EIS-0245-SA-03: Supplement Analysis Management of Spent Nuclear Fuel from the K Basins at the Hanford Site, Richland, Washington August 1, 2001 EIS-0245-SA-02: Supplement Analysis Management of Spent Nuclear Fuel from the K Basins at the Hanford Site, Richland, Washington August 1, 1998 EIS-0245-SA-01: Supplement Analysis Management of Spent Nuclear Fuel from the K Basins at the Hanford Site,

155

EM Safely and Efficiently Manages Spent Nuclear Fuel | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Waste Management » Nuclear Materials & Waste » EM Services » Waste Management » Nuclear Materials & Waste » EM Safely and Efficiently Manages Spent Nuclear Fuel EM Safely and Efficiently Manages Spent Nuclear Fuel Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. EM's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal in a geologic repository. Previously, the Office of Environmental Management's (EM) mission had included the safe and efficient management of its spent nuclear fuel (SNF) and preparation for its disposal in a geologic repository. However, in May 2009, the planned geologic repository at Yucca Mountain was cancelled. The

156

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

National Report Joint Convention on the Safety of Spent Fuel 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 referred to as the "Joint Convention". This report satisfies the requirements of the Joint Convention for reporting on the status of safety at spent fuel and radioactive waste management facilities within the United States of America (U.S.). National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management - May 2003

157

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

SciTech Connect

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.

Rodney C. Ewing

2004-10-07T23:59:59.000Z

158

Test protocol for aluminum based spent nuclear fuel  

SciTech Connect

Aluminum based spent nuclear fuel (Al-SNF) will be treated and then stored in a geological repository. A series of corrosion tests, including coupon, galvanic, and electrochemical, were performed to measure the effects of degradation on U-Al alloys, which simulated various Al-SNF forms. The test solutions were variants of a J-13 well water chemistry. As part of the degradation characteristics, the dissolution of uranium, which is the primary radionuclide, was measured. The degradation and dissolution were changed by the type of water chemistry and the temperature. The alloy composition and fabrication were not significant variables.

Mickalonis, J.I.; Wiersma, B.J.

1999-07-01T23:59:59.000Z

159

Closure mechanism and method for spent nuclear fuel canisters  

DOE Patents (OSTI)

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.

Doman, Marvin J. (Monroeville, PA)

2004-11-23T23:59:59.000Z

160

Production of metal waste forms from spent fuel treatment  

SciTech Connect

Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities.

Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.

1995-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

FOREIGN ASSISTANCE FOR HANDLING SPENT NUCLEAR NAVAL FUEL IN RUSSIA: SETTING PRIORITIESi  

Science Journals Connector (OSTI)

For Russia's foreign partners, prioritizing tasks related to spent fuel management first involves identifying urgent needs, potential bottlenecks or gaps in assistance programs, safety and security issues, and...

C. CHUEN

2006-01-01T23:59:59.000Z

162

Instrumented, Shielded Test Canister System for Evaluation of Spent Nuclear Fuel in Dry Storage  

SciTech Connect

This document describes the development of an instrumented, shielded test canister system to store and monitor aluminum-based spent nuclear duel under dry storage conditions.

Sindelar, R.L.

1999-10-21T23:59:59.000Z

163

E-Print Network 3.0 - advanced spent fuel Sample Search Results  

NLE Websites -- All DOE Office Websites (Extended Search)

facilities should have more advanced technical monitoring... of the cycle to extract fis- sile material from the spent fuel removed from reactors. Although a complete... of...

164

Behavior of Spent Nuclear Fuel in Water Pool Storage  

Office of Scientific and Technical Information (OSTI)

Behavior of Spent Nuclear Behavior of Spent Nuclear Fuel in Water Pool Storage A. 0; Johnson, jr. , I ..: . Prepared Cor the Energy Research and Development Administration under Contract EY-76-C-06-1830 ---- Pat t i ~ < N ~ ~ r ~ t b w t ~ - ! I , ~ I ~ ~ ~ I . I I ~ ) ~ I I ~ ~ N O T I C E T€& - was prepad pnpn4. m w n t of w k spon-d by the Unitd S t . & ) C a u n m ~ (*WU ij*. M t e d $tam w the Wqy R e s e w & a d Ohrsropmcnt ~dmhirmlion, nor m y d thair ewhew,,nq Pny @fw a n t r ~ ~ t 0 ~ 1 , s ~ k m r i t r i l t t q r , ~ , m r tWf ernpfQw, r(tLltm any wartany, s x p r e s or kWld,= w w aAql -9 . o r r w p a m l ~ ~ t y for e~ o r uodruincvr of any infomutim, 9 F p d + d - , or repratants that -would nat 1 d - e privately owned rfghas. ,i PAQFIC NORTHWEST UBORATORY operated b ;"' SArnLLE ' fw the E M R m RESEARCH AND DEVELOPMENT ADMINISTRAT1QN Wk.Cwfraa rv-76c-ts-is38

165

Fracture behavior of high-burnup spent-fuel cladding  

SciTech Connect

PCI-like, brittle-type failures, characterized by pseudocleavage-plus-fluting features in the fracture surface, branching cracks, and small diametral strain, were observed to occur at 292 to 325/sup 0/C in some batches of spent power-reactor fuel-cladding tubes under internal gas-pressurization and expanding-mandrel loading conditions in which the tests were not influenced by fission product simulants. Fractographic characteristics per se do not provide evidence for a PCI failure mechanism but should be deemed only as cooroborative in nature. Evaluation of TEM thin-foil specimens, obtained from regions adjacent to the brittle-type fracture sites, characteristically revealed extensive amounts of Zr/sub 3/O precipitates and a lack of slip dislocations. The precipitation of the Zr/sub 3/O phase appears to be enhanced by a high density of irradiation-induced defects. The brittle-type failure produced in the spent-fuel cladding tubes appears to be associated with segregation of oxygen to dislocation substructures and irradiation-induced defects, which leads to the formation of an ordered zirconium-oxygen phase of Zr/sub 3/O, an immobilization of dislocations, and minimal plastic deformation in the cladding material.

Chung, H.M.; Yaggee, F.L.; Kassner, T.F.

1983-10-01T23:59:59.000Z

166

Drop Testing of DOE Spent Nuclear Fuel Canisters  

SciTech Connect

The National Spent Nuclear Fuel Program (NSNFP) at the Idaho National Engineering and Environmental Laboratory INEEL) prepared four representative Department of Energy DOE) spent nuclear fuel (SNF) canisters for the purpose of drop testing. The first two canisters represented a modified 24- inch diameter standardized DOE SNF canister and the second two canisters represented the Hanford Multi-Canister Overpack MCO). The modified canisters and internals were constructed and assembled at the INEEL. The MCO internal weights were fabricated at the INEEL and assembled into two MCOs at Hanford and later shipped to the INEEL for drop test preparation. Drop testing of these four canisters was completed in August 2004 at Sandia National Laboratories. The modified canisters were dropped from 30 feet onto a flat, essentially unyielding surface, with the canisters oriented at 45 degrees and 70 degrees off-vertical at impact. One representative MCO was dropped from 23 feet onto the same flat surface, oriented vertically at impact. The second representative MCO was dropped onto the flat surface from 2 feet oriented at 60 degrees off-vertical. These drop heights and orientations were chosen to meet or exceed the Yucca Mountain repository drop criteria. This paper discusses the comparison of deformations between the actual dropped canisters and those predicted by pre-drop and limited post-drop finite element evaluations performed using ABAQUS/Explicit. Post-drop containment of all four canisters, demonstrated by way of helium leak testing, is also discussed.

S. D. Snow; D. K. Morton; T. E. Rahl; R. K. Blandford; T. J. Hill

2005-07-01T23:59:59.000Z

167

Spent Fuel Storage Operational Experience With Increased Crud Activities  

SciTech Connect

A significant part of the electricity production in Hungary is provided by 4 units of VVER 440 nuclear reactors at the Paks Nuclear Power Plant. Interim dry storage of the spent fuel assemblies that are generated during the operation of the reactors is provided in a Modular Vault Dry Storage (MVDS) facility that is located in the immediate vicinity of the Paks Nuclear Power Plant. The storage capacity of the MVDS is being continuously extended in accordance with spent the fuel production rate from the four reactors. An accident occurred at unit 2 of the Paks Nuclear Power Plant in 2003, when thirty irradiated fuel assemblies were damaged during a cleaning process. The fuel assemblies were not inside the reactor at the time of the accident, but in a separate tank within the adjacent fuel decay pool. As a result of this accident, contamination from the badly damaged fuel assemblies spread to the decay pool water and also became deposited onto the surface of (hermetic) spent fuel assemblies within the decay pool. Therefore, it was necessary to review the design basis of the MVDS and assess the effects of taking the surface contaminated spent fuel assemblies into dry storage. The contaminated hermetic assemblies were transferred from the unit 2 pool to the interim storage facility in the period between 2005 and 2007. Continuous inspection and measurement was carried out during the transfer of these fuel assemblies. On the basis of the design assessments and measurement of the results during the fuel transfer, it was shown that radiological activity values increased due to the consequences of the accident but that these levels did not compromise the release and radiation dose limits for the storage facility. The aim of this paper is to show the effect on the operation of the MVDS interim storage facility as a result of the increased activity values due to the accident that occurred in 2003, as well as to describe the measurements that were taken, and their results and experience gained. In summary: On the basis of the design assessments and measurement of the results during the fuel transfer operations, it was shown that radiological activity values increased due to the consequences of the 2003 accident but that these levels did not compromise the release and dose limits for the fuel storage facility. In the environment there was no measurable radioactivity as a result of the operation of the Paks ISFSI. The exposure of the surrounding population was calculated on measured releases and meteorological data. The calculations show negligible doses until 2004. Due to the increased surface contamination on the spent fuel assemblies the dose rate increased almost 5 times compared to the least annual value, but still less then 0.01 percent of the allowed dose restriction. (authors)

Barnabas, I. [Public Agency for Radioactive Waste, Management (PURAM) (Hungary); Eigner, T. [Paks NPP (Hungary); Gresits, I. [Technical University of Budapest (Hungary); Ordagh, M. [SOM System Llc, (Hungary)

2008-07-01T23:59:59.000Z

168

DECONTAMINATION OF ZIRCALOY CLADDING HULLS FROM SPENT NUCLEAR FUEL  

SciTech Connect

The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The {sup 90}Sr and {sup 137}Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 {micro}m of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit. Reprocessing of spent nuclear fuel generates a cladding waste which would likely require disposal as a Greater than Class C LLW in the US. If the cladding hulls could be treated to remove a majority of the actinide and fission product contamination, the hulls could potentially meet acceptance criteria for disposal as a LLW or allow recycle of the Zr metal. Discard of the hulls as a LLW would result in significant cost savings compared to disposal as a Greater than Class C waste which currently has no disposition path. During fuel irradiation and reprocessing, radioactive materials are produced and deposited in the Zircaloy cladding. Due to short depths of penetration, the majority of the fission products and actinide elements are located in the ZrO{sub 2} layer which forms on the surface of the cladding during fuel irradiation. Therefore, if the oxide layer is removed, the majority of the contamination should also be removed. It is very difficult, if not impossible to remove all of the activity from spent fuel cladding since traces of U and Th in the unirradiated Zircaloy adsorb neutrons generating higher actinides in the bulk material. During fuel irradiation, {sup 92}Zr is also converted to radioactive {sup 93}Zr by neutron adsorption. Methods for decontaminating and conditioning irradiated Zircaloy cladding hulls have been investigated in Europe, Japan, and the US during the last 35 years; however, a method to decontaminate the hulls to an activity level which meets US acceptance criteria for disposal as a LLW was not deployed on a commercial scale. The feasibility of decontaminating spent fuel cladding hulls was investigated as part of the GNEP Separations Campaign. Small-scale experiments were used to demonstrate the removal of the ZrO{sub 2} layer from Zircaloy coupons using dilute solutions ({le}1.0 M) of HF. 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. To test the HF decontamination process, experiments were subsequently performed using actual spent fuel cladding hulls. Decontamination experiments were performed to measure the fission product and actinide concentrations as a function of the depth of the surface removed from the cladding hull. The experimental methods used to perform these experiments and a discussion of the results and observations are presented in the following sections.

Rudisill, T.

2010-09-29T23:59:59.000Z

169

International Atomic Energy Agency (IAEA) activities on spent fuel management options  

SciTech Connect

Many countries have in the past several decades opted for storage of spent fuel for undefined periods of time. They have adopted the 'wait and see' strategy for spent fuel management. A relatively small number of countries have adopted reprocessing and use of MOX fuel as part of their strategy in spent fuel management. From the 10, 000 tonnes of heavy metal that is removed annually from nuclear reactors throughout the world, only approximately 30 % is currently being reprocessed. Continuous re-evaluation of world energy resources, announcement of the Global Nuclear Energy Partnership (GNEP) and the Russian initiative to form international nuclear centers, including reprocessing, are changing the stage for future development of nuclear energy. World energy demand is expected to more than double by 2050, and expansion of nuclear energy is a key to meeting this demand while reducing pollution and greenhouse gases. Since its foundation, the International Atomic Energy Agency (IAEA) has served as an interface between countries in exchanging information on the peaceful development of nuclear energy and at the same time guarding against proliferation of materials that could be used for nuclear weapons. The IAEA's Department of Nuclear Energy has been generating technical documents, holding meetings and conferences, and supporting technical cooperation projects to facilitate this exchange of information. This paper focuses on the current status of IAEA activities in the field of spent fuel management being carried out by the Division of Nuclear Fuel Cycle and Waste Technology. Information on those activities could be found on the web site link www.iaea.org/OurWork/ST/NE/NEFW/nfcms. To date, the IAEA has given priority in its spent fuel management activities to supporting Member States in their efforts to deal with growing accumulations of spent power reactor fuel. There is technical consensus that the present technologies for spent fuel storage, wet and dry, provide adequate protection to people and environment. As storage durations grow, the IAEA has expanded its work related to the implications of extended storage periods. Operation and maintenance of containers for storage and transport have also been investigated related to long term storage periods. In addition, as international interest in reprocessing of spent fuel increases, the IAEA continues to serve as a crossroads for sharing the latest developments in spent fuel treatment options. A Coordinated Research Project is currently addressing spent fuel performance assessment and research to evaluate long term effects of storage on spent fuel. The effect of increased burnup and mixed oxide fuels on spent fuel management is also the focus of interest as it follows the trend in optimizing the use of nuclear fuel. Implications of damaged fuel on storage and transport as well as burnup credit in spent fuel applications are areas that the IAEA is also investigating. Since spent fuel management considerations require social stability and institutional control, those aspects are taken into account in most IAEA activities. Data requirements and records management as storage durations extend were also investigated as well as the potential for regional spent fuel storage facilities. Spent fuel management activities continue to be coordinated with others in the IAEA to ensure compliance and consistency with efforts in the Department of Safety and Security and the Department of Safeguards, as well as with activities related to geologic disposal. Either disposal of radioactive waste or spent fuel will be an ultimate consideration in all spent fuel management options. Updated information on spent fuel treatment options that include fuel reprocessing as well as transmutation of minor actinides are investigated to optimize the use of nuclear fuel and minimize impact on environment. Tools for spent fuel management economics are also investigated to facilitate assessment of industrial applicability for these options. Most IAEA spent fuel management activities will ultimately be reported in o

Lovasic, Z.; Danker, W. [International Atomic Energy Agency (IAEA) Vienna (Austria)

2007-07-01T23:59:59.000Z

170

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

171

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

J.K. Knudson

2003-10-02T23:59:59.000Z

172

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Second National Report for the Joint Convention on the Safety of 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 Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention). This report satisfies the requirements of the Joint Convention for reporting on the status of safety at spent fuel (SF) and radioactive waste management facilities within the United States of America (U.S.). Second National Report for the Joint Convention on the Safety of Spent Fuel

173

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

Gasoline and Diesel Fuel Update (EIA)

Home > Nuclear > Spent Nuclear Fuel Home > Nuclear > Spent Nuclear Fuel Release Date: October 1, 2004 Next Release: Late 2010** Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive Waste Management (OCRWM). The spent nuclear fuel (SNF) data includes detailed characteristics of SNF generated by commercial U.S. nuclear power plants. From 1983 through 1995 this data was collected annually. Since 1996 this data has been collected every three years. The latest available detailed data covers all SNF discharged from commercial reactors before December 31, 2002, and is maintained in a data base by the EIA. Summary data tables from this data base may be found as indicated below. Table 1. Total U.S. Commercial Spent Nuclear Fuel Discharges, 1968 - 2002

174

Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel  

SciTech Connect

Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

2004-12-27T23:59:59.000Z

175

Management of super-grade plutonium in spent nuclear fuel  

SciTech Connect

This paper examines the security and safeguards implications of potential management options for DOE's sodium-bonded blanket fuel from the EBR-II and the Fermi-1 fast reactors. The EBR-II fuel appears to be unsuitable for the packaging alternative because of DOE's current safeguards requirements for plutonium. Emerging DOE requirements, National Academy of Sciences recommendations, draft waste acceptance requirements for Yucca Mountain and IAEA requirements for similar fuel also emphasize the importance of safeguards in spent fuel management. Electrometallurgical treatment would be acceptable for both fuel types. Meeting the known requirements for safeguards and security could potentially add more than $200M in cost to the packaging option for the EBR-II fuel.

McFarlane, H. F.; Benedict, R. W.

2000-03-20T23:59:59.000Z

176

SPENT FUEL MANAGEMENT AT THE SAVANNAH RIVER SITE  

SciTech Connect

Spent nuclear fuels are received from reactor sites around the world and are being stored in the L-Basin at the Savannah River Site (SRS) in Aiken, South Carolina. The predominant fuel types are research reactor fuel with aluminum-alloy cladding and aluminum-based fuel. Other fuel materials include stainless steel and Zircaloy cladding with uranium oxide fuel. Chemistry control and corrosion surveillance programs have been established and upgraded since the early 1990's to minimize corrosion degradation of the aluminum cladding materials, so as to maintain fuel integrity and minimize personnel exposure from radioactivity in the basin water. Recent activities have been initiated to support additional decades of wet storage which include fuel inspection and corrosion testing to evaluate the effects of specific water impurity species on corrosion attack.

Vormelker, P; Robert Sindelar, R; Richard Deible, R

2007-11-03T23:59:59.000Z

177

Spent nuclear fuel policies in historical perspective: An international comparison  

Science Journals Connector (OSTI)

The purpose of this article is to explain why the world's nuclear power countries differ from each other with respect to their spent nuclear fuel (SNF) policies. The emergence and evolution of three principal SNF approaches are analyzed: direct disposal, reprocessing and SNF export. Five broad explanatory factors are identified and discussed in relation to the observed differences in policy outcomes: military ambitions and non-proliferation, technological culture, political culture and civil society, geological conditions, and energy policy. SNF policy outcomes can generally be seen to result from a complex interaction between these broad factors, but it is also possible to discern a number of important patterns. To the extent that the five factors may undergo far-reaching changes in the future, the historical experience of how they have shaped SNF policies also give a hint of possible future directions in SNF policymaking around the world.

Per Högselius

2009-01-01T23:59:59.000Z

178

Structural analysis of extracts from spent hydroprocessing catalysts  

SciTech Connect

Solvent extracts from spent commercial naphtha and gas oil hydrotreating catalysts were examined in order to elucidate the structure of adsorbed poisons. Amides were identified in the extracts and made up 20-30% of the basic compounds. The oxygen content of the extracts was as high as 28% by weight. Some of this oxygen may be associated with metal complexes or present as inorganic material. Infrared spectroscopic analysis revealed very strong absorptions due to carbonyls, suggesting that compounds such as ketones and carboxylic acids were present. The extracted compounds consisted of saturated (paraffinic and naphthenic) structures and aromatic groups, with >50% boiling at +343/sup 0/c. These results suggest polymerization of species on the surface of the catalyst, rather than condensation of aromatic structures to form coke.

Choi, J.H.K.; Gray, M.R.

1988-09-01T23:59:59.000Z

179

Training implementation matrix, Spent Nuclear Fuel Project (SNFP)  

SciTech Connect

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.

EATON, G.L.

2000-06-08T23:59:59.000Z

180

Fracture behavior of zircaloy spent-fuel cladding  

SciTech Connect

The Zircaloy cladding of water reactor fuel rods is susceptible to local breach-type failure, commonly known as pellet-cladding interaction (PCI) failure, during operational and off-normal power transients after the fuel has achieved a sufficiently high burnup. An optimization of power ramp procedures or fuel rod fabrication to minimize the cladding failure would result in a significant decrease in radiation exposure of plant personnel due to background and airborne radioactivity as well as an extension of core life in terms of allowable off-gas radioactivity. As part of a program to provide a better understanding of the fuel rod faiure phenomenon and to facilitate the formulation of a better failure criterion, a mechanistic study of the deformation and fracture behavior of high-burnup spent-fuel cladding is in progress under simulated PCI conditions.

Chung, H.M.; Yaggee, F.L.; Kassner, T.F.

1983-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
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181

LCA of a spent lube oil Re-refining process  

Science Journals Connector (OSTI)

Although re-refining of spent lubricating oils (used oils) has been practiced with varying technical and commercial success for over the past 50 years, a sustainable processing technology has yet to become widely accepted. Poor on-stream efficiency, inconsistent product quality, and careless management of feedstock contaminants and byproducts have often resulted in widespread environmental problems and poor economics. Environmentally-conscious design of processes and products is increasingly viewed as an integral strategy in the sustainable development of new refining and chemical processes. Life cycle assessment is becoming the preferred methodology for comparing the environmental impacts of competing processes. A life cycle analyses of a promising new re-refining technology, the HyLubeTM process, has been undertaken to quantify the intrinsic benefits of HyLube re-refining over the current practice of recovering used oils for fuel value

Tom N. Kalnes; David R. Shonnard; Andreas Schuppel

2006-01-01T23:59:59.000Z

182

Reverse logistics network design for spent batteries: a simulation study  

Science Journals Connector (OSTI)

End of life (EOL) product management, which encompasses reuse, remanufacturing and materials recycling, requires a structured reverse logistic network in order to collect products efficiently at the end of their life cycle. This work describes modelling and simulation of reverse logistics network design for collection of spent batteries for Sangrur District of North India. To compare different order assignment, a simulation model of forward and reverse logistics networks has been developed. Several simulation experiments have been designed to analyse impact of the system design factors on the operational performance of the reverse logistics system. The simulation results show that the model presented in this paper calculates the battery collection cost, transfer time, transfer cost, and resource utilisation in a predictable manner. Moreover, it provides a tool to understand how the system behaves by carrying out 'what-if' assessments and to identify which parameters are most important for more detailed analysis.

Arvind Jayant; Pardeep Gupta; S.K. Garg

2014-01-01T23:59:59.000Z

183

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

184

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

SciTech Connect

In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of determining Pu mass in spent fuel assemblies.

Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

2011-06-30T23:59:59.000Z

185

Deformation and fracture characteristics of spent Zircaloy fuel cladding  

SciTech Connect

For a better understanding of Zircaloy fuel-rod failure by the pellet-cladding interaction (PCI) phenomenon, a mechanistic study of deformation and fracture behavior of spent power reactor fuel cladding under simulated PCI conditions was conducted. Zircaloy-2 cladding specimens, obtained from fuel assemblies of operating power reactors, were deformed to fracture at 325/sup 0/C by internal gas pressurization in the absence of fission product simulants. Fracture characteristics and microstructures were examined via SEM, TEM, and HVEM. Numerous dislocation tangles and cell structures, observed in TEM specimens of cladding tubes that failed in a ductile manner, were consistent with SEM observations of a limited number of dimples characteristic of microvoid coalescence. A number of brittle-type failures were produced without the influence of fission product simulants. The brittle cracks occurred near the areas compressed by the Swagelok fittings of the internally pressurized tube and propagated from the outer to the inner surface. Since the outer surface was isolated and maintained under a flowing stream of pure helium, it is unlikely that the brittle-type failure was influenced by any fission product traces. SEM fractography of the brittle-type failure revealed a large area of transgranular pseudocleavage with limited areas of ductile fluting, which were similar in appearance to the surfaces produced by in-reactor PCI-type failures. A TEM evaluation of the cladding in the vicinity of the through-wall crack revealed numerous locations that contained an extensive amount of second-phase precipitate (Zr/sub 3/O). We believe that the brittle-type failures of the irradiated spent fuel cladding in the stress rupture experiments are associated with segregation of oxygen, which leads to the formation of the order structure, an immobilization of dislocations, and minimal plastic deformation in the material.

Chung, H.M.; Yaggee, F.L.

1982-09-01T23:59:59.000Z

186

Issues related to EM management of DOE spent nuclear fuel  

SciTech Connect

This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE`s SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE`s national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references.

Abbott, D.G. [EG& G Idaho, Inc., Idaho Falls, ID (United States); Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M. [IT Corp. (United States)

1993-07-01T23:59:59.000Z

187

A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite  

E-Print Network (OSTI)

geological disposal of spent CANDU fuel in Canada, a safetyhypothetical repository for spent CANDU fuel in the Canadianbuffer. The waste form: CANDU reactors in Canada are fuelled

Nguyen, T.S.

2009-01-01T23:59:59.000Z

188

Spent Nuclear Fuel (SNF) Project Acceptance Criteria for Light Water Reactor Spent Fuel Storage System [OCRWM PER REV2  

SciTech Connect

As part of the decommissioning of the 324 Building Radiochemical Engineering Cells there is a need to remove commercial Light Water Reactor (LWR) spent nuclear fuel (SNF) presently stored in these hot cells. To enable fuel removal from the hot cells, the commercial LWR SNF will be packaged and shipped to the 200 Area Interim Storage Area (ISA) in a manner that satisfies site requirements for SNF interim storage. This document identifies the criteria that the 324 Building Radiochemical Engineering Cell Clean-out Project must satisfy for acceptance of the LWR SNF by the SNF Project at the 200 Area ISA. In addition to the acceptance criteria identified herein, acceptance is contingent on adherence to applicable Project Hanford Management Contract requirements and procedures in place at the time of work execution.

JOHNSON, D.M.

2000-12-20T23:59:59.000Z

189

In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies  

SciTech Connect

A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

2008-04-16T23:59:59.000Z

190

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

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.

Not Available

1994-06-01T23:59:59.000Z

191

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321  

SciTech Connect

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

2013-07-01T23:59:59.000Z

192

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

SciTech Connect

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).

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

1982-09-01T23:59:59.000Z

193

Accelerator-driven transmutation of spent fuel elements  

DOE Patents (OSTI)

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

Venneri, Francesco (Los Alamos, NM); Williamson, Mark A. (Los Alamos, NM); Li, Ning (Los Alamos, NM)

2002-01-01T23:59:59.000Z

194

Interim Storage of Hanford Spent Fuel & Associated Sludge  

SciTech Connect

The Hanford site is currently dealing with a number of types of Spent Nuclear Fuel. The route to interim dry storage for the various fuel types branches along two different paths. Fuel types such as metallic N reactor fuel and Shippingport Core 2 Blanket assemblies are being placed in approximately 4 m long canisters which are then stored in tubes below grade in a new canister storage building. Other fuels such as TRIGA{trademark} and Light Water Reactor fuel will be relocated and stored in stand-alone casks on a concrete pad. Varying degrees of sophistication are being applied with respect to the drying and/or evacuation of the fuel interim storage canisters depending on the reactivity of the fuel, the degree of damaged fuel and the previous storage environment. The characterization of sludge from the Hanford K Basins is nearly complete and canisters are being designed to store the sludge (including uranium particles from fuel element cleaning) on an interim basis.

MAKENAS, B.J.

2002-07-01T23:59:59.000Z

195

Radionuclide release rates from spent fuel for performance assessment modeling  

SciTech Connect

In a scenario of aqueous transport from a high-level radioactive waste repository, the concentration of radionuclides in water in contact with the waste constitutes the source term for transport models, and as such represents a fundamental component of all performance assessment models. Many laboratory experiments have been done to characterize release rates and understand processes influencing radionuclide release rates from irradiated nuclear fuel. Natural analogues of these waste forms have been studied to obtain information regarding the long-term stability of potential waste forms in complex natural systems. This information from diverse sources must be brought together to develop and defend methods used to define source terms for performance assessment models. In this manuscript examples of measures of radionuclide release rates from spent nuclear fuel or analogues of nuclear fuel are presented. Each example represents a very different approach to obtaining a numerical measure and each has its limitations. There is no way to obtain an unambiguous measure of this or any parameter used in performance assessment codes for evaluating the effects of processes operative over many millennia. The examples are intended to suggest by example that in the absence of the ability to evaluate accuracy and precision, consistency of a broadly based set of data can be used as circumstantial evidence to defend the choice of parameters used in performance assessments.

Curtis, D.B.

1994-11-01T23:59:59.000Z

196

Refinishing contamination floors in Spent Nuclear Fuels storage basins  

SciTech Connect

The floors of the K Basins at the Hanford Site are refinished to make decontamination easier if spills occur as the spent nuclear fuel (SNF) is being unloaded from the basins for shipment to dry storage. Without removing the contaminated existing coating, the basin floors are to be coated with an epoxy coating material selected on the basis of the results of field tests of several paint products. The floor refinishing activities must be reviewed by a management review board to ensure that work can be performed in a controlled manner. Major documents prepared for management board review include a report on maintaining radiation exposure as low as reasonably achievable, a waste management plan, and reports on hazard classification and unreviewed safety questions. To protect personnel working in the radiation zone, Operational Health Physics prescribed the required minimum protective methods and devices in the radiological work permit. Also, industrial hygiene safety must be analyzed to establish respirator requirements for persons working in the basins. The procedure and requirements for the refinishing work are detailed in a work package approved by all safety engineers. After the refinishing work is completed, waste materials generated from the refinishing work must be disposed of according to the waste management plan.

Huang, F.F.; Moore, F.W.

1997-07-11T23:59:59.000Z

197

Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability  

SciTech Connect

This paper examines long-term cost differences between two competing Light Water Reactor (LWR) fuels: Uranium Oxide (UOX) and Mixed Uranium Oxide-Plutonium Oxide (MOX). Since these costs are calculated on a life-cycle basis, expected savings from lower future MOX fuel prices can be used to value the option of substituting MOX for UOX, including the value of maintaining access to the used UOX fuel that could be reprocessed to make MOX. The two most influential cost drivers are the price of natural uranium and the cost of reprocessing. Significant and sustained reductions in reprocessing costs and/or sustained increases in uranium prices are required to give positive value to the retrievability of Spent Nuclear Fuel. While this option has positive economic value, it might not be exercised for 50 to 200 years. Therefore, there are many years for a program during which reprocessing technology can be researched, developed, demonstrated, and deployed. Further research is required to determine whether the cost of such a program would yield positive net present value and/or increases the sustainability of LWR energy systems.

Wood, Thomas W.; Rothwell, Geoffrey

2012-01-10T23:59:59.000Z

198

Multi-Detector Analysis System for Spent Nuclear Fuel Characterization  

SciTech Connect

The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

1999-09-01T23:59:59.000Z

199

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fourth National Report for the Joint Convention on the Safety of Fourth National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management Fourth National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management 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 status of safety at spent fuel and radioactive waste management facilities within the U.S.

200

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Third National Report for the Joint Convention on the Safety of 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 Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management 1(Joint Convention). This report reflects developments in the United States through September 2008. This report satisfies the requirements of the Joint Convention for reporting on the status of safety at spent fuel and radioactive waste management facilities within

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

6 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 (storage and transport) casks. ii Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel EXECUTIVE SUMMARY The House Appropriations Committee Print that accompanied the Consolidated Appropriations Act, 2008, requests that the U.S. Department of Energy (the Department):

202

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada Summary 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 radioactive waste. The EIS evaluates not only impacts from constructing, operating, monitoring, and closing a repository, but also from transporting the materials from 72 commercial and 4 DOE sites to the Yucca Mountain repository site in Nye County, Nevada. Public Comment Opportunities

203

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

Energy.gov (U.S. Department of Energy (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.

204

Stainless steel-zirconium waste forms from the treatment of spent nuclear fuel  

Science Journals Connector (OSTI)

Stainless steel-zirconium waste-form alloys have been developed for the disposal of metallic wastes recovered from spent nuclear fuel using the electrometallurgical process developed by Argonne National Laborator...

S. M. McDeavitt; D. P. Abraham; J. Y. Park; D. D. Keiser

1997-07-01T23:59:59.000Z

205

IF GEORGE BUSH spent more time and money on mobilising Weapons of Mass  

E-Print Network (OSTI)

IF GEORGE BUSH spent more time and money on mobilising Weapons of Mass Salvation (WMS) in addition to combating Weapons of Mass Destruction (WMD), we might actually get somewhere in making this planet a safer

206

Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities  

SciTech Connect

This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

MITCHELL, R.M.

2000-09-28T23:59:59.000Z

207

Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)  

Energy.gov (U.S. Department of Energy (DOE))

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...

208

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

Energy.gov (U.S. Department of Energy (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...

209

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

SciTech Connect

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.

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

2014-06-10T23:59:59.000Z

210

Spent Shale Grouting of Abandoned In-Situ Oil Shale Retorts  

E-Print Network (OSTI)

for the grout. SPENT SHALE Oil shale, which is a low-gradeMineral Reactions in Colorado Oil Shale," Lawrence Livermore1978. of Decomposition of Colorado Oil Shale: II. Livermore

Fox, J.P.; Persoff, P.

1980-01-01T23:59:59.000Z

211

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

DOE Patents (OSTI)

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.

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

2013-11-05T23:59:59.000Z

212

Spent nuclear fuel and residential property values: the influence of proximity, visual cues and public information  

Science Journals Connector (OSTI)

...This article examines whether public knowledge of spent fuel storage at nuclear power plants, and any ... that may have occurred, affect the sale price of single-family residential properties. We present ... m...

David E. Clark; Tim Allison

1999-10-01T23:59:59.000Z

213

Plasma processing of spent nuclear fuel by two-frequency ion cyclotron resonance heating  

Science Journals Connector (OSTI)

A previously developed method for analyzing the plasma processing of spent nuclear fuel is generalized to a plasma containing multicharged fuel ions. In such a plasma, ion cyclotron resonance heating of nuclear a...

A. V. Timofeev

2009-11-01T23:59:59.000Z

214

Optimization of ethanol production from spent tea waste by Saccharomyces cerevisiae using statistical experimental designs  

Science Journals Connector (OSTI)

The aim of this study was to investigate the prospect for the use of spent tea waste (STW), an important municipal waste, as a potential substrate to generate hydrolysates for fuel ethanol production. Acid pretre...

Yasin Yücel; Sezer Göyc?nc?k

2014-07-01T23:59:59.000Z

215

Plasma processing of spent nuclear fuel by two-frequency ion cyclotron resonance heating  

SciTech Connect

A previously developed method for analyzing the plasma processing of spent nuclear fuel is generalized to a plasma containing multicharged fuel ions. In such a plasma, ion cyclotron resonance heating of nuclear ash ions should be carried out in two monochromatic RF fields of different frequencies, provided that the fraction of {xi} multicharged ions is small, {xi} {<=} 0.1, a condition that substantially restricts the productivity of systems for processing spent nuclear fuel. Ways of overcoming this difficulty are discussed.

Timofeev, A. V. [Russian Research Centre Kurchatov Institute, Nuclear Fusion Institute (Russian Federation)

2009-11-15T23:59:59.000Z

216

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

SciTech Connect

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.

Bailey, W.J.

1990-02-01T23:59:59.000Z

217

Estimation of the Transportation Risks for the Spent Fuel in Korea for Various Transportation Scenarios  

SciTech Connect

According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. Also, we estimated and compared the transportation risks for these four transportation scenarios. From the results of this study, we found that these four transportation scenarios for spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future. (authors)

Jongtae, Jeong; Cho, D.K.; Choi, H.J.; Choi, J.W. [Korea Atomic Energy Research Institute, Yuseong, Daejeon (Korea, Republic of)

2008-07-01T23:59:59.000Z

218

Achieving increased spent fuel storage capacity at the High Flux Isotope Reactor (HFIR)  

SciTech Connect

The HFIR facility was originally designed to store approximately 25 spent cores, sufficient to allow for operational contingencies and for cooling prior to off-site shipment for reprocessing. The original capacity has now been increased to 60 positions, of which 53 are currently filled (September 1994). Additional spent cores are produced at a rate of about 10 or 11 per year. Continued HFIR operation, therefore, depends on a significant near-term expansion of the pool storage capacity, as well as on a future capability of reprocessing or other storage alternatives once the practical capacity of the pool is reached. To store the much larger inventory of spent fuel that may remain on-site under various future scenarios, the pool capacity is being increased in a phased manner through installation of a new multi-tier spent fuel rack design for higher density storage. A total of 143 positions was used for this paper as the maximum practical pool capacity without impacting operations; however, greater ultimate capacities were addressed in the supporting analyses and approval documents. This paper addresses issues related to the pool storage expansion including (1) seismic effects on the three-tier storage arrays, (2) thermal performance of the new arrays, (3) spent fuel cladding corrosion concerns related to the longer period of pool storage, and (4) impacts of increased spent fuel inventory on the pool water quality, water treatment systems, and LLLW volume.

Cook, D.H.; Chang, S.J.; Dabs, R.D.; Freels, J.D.; Morgan, K.A.; Rothrock, R.B. [Oak Ridge National Lab., TN (United States); Griess, J.C. [Griess (J.C.), Knoxville, TN (United States)

1994-12-31T23:59:59.000Z

219

Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel  

SciTech Connect

Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed.

Dahl, C.A.; Adams, J.P.; Ramer, R.J.

1998-07-01T23:59:59.000Z

220

Technology development program for Idaho Chemical Processing Plant spent fuel and waste management  

SciTech Connect

Acidic high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the U.S. Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel 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, and describes the Spent Fuel and HLW Technology program in more detail.

Ermold, L.F.; Knecht, D.A.; Hogg, G.W.; Olson, A.L.

1993-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

International auspices for the storage of spent nuclear fuel as a nonproliferation measure  

SciTech Connect

The maintenance of spent nuclear fuel from power reactors will pose problems regardless of how or when the debate over reprocessing is resolved. At present, many reactor sites contain significant buildups of spent fuel stored in holding pools, and no measure short of shutting down reactors with no remaining storage capacity will alleviate the need for away-from-reactor storage. Although the federal government has committed itself to dealing with the spent fuel problem, no solution has been reached, largely because of a debate over differing projections of storage capacity requirements. Proliferation of weapons grade nuclear material in many nations presents another pressing issue. If nations with small nuclear programs are forced to deal with their own spent fuel accumulations, they will either have to reprocess it indigenously or contract to have it reprocessed in a foreign reprocessing plant. In either case, these nations may eventually possess sufficient resources to assemble a nuclear weapon. The problem of spent fuel management demands real global solutions, and further delay in solving the problem of spent nuclear fuel accumulation, both nationally and globally, can benefit only a small class of elected officials in the short term and may inflict substantial costs on the American public, and possibly the world. (JMT)

O'Brien, J.N.

1981-10-01T23:59:59.000Z

222

Uncanistered Spent Nuclear fuel Disposal Container System Description Document  

SciTech Connect

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.

NONE

2000-10-12T23:59:59.000Z

223

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

SciTech Connect

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 country’s 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.

Trond Bjornard; Philip C. Durst

2012-05-01T23:59:59.000Z

224

COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel  

SciTech Connect

Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

1986-12-01T23:59:59.000Z

225

Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.  

SciTech Connect

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 data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This 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 the Nuclear Regulatory Commission.

Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

2005-07-01T23:59:59.000Z

226

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology  

SciTech Connect

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which has the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu; Kinoshita, Kensuke

2005-05-15T23:59:59.000Z

227

5.14 - Spent Fuel Dissolution and Reprocessing Processes  

Science Journals Connector (OSTI)

Abstract The initial motivation for the development of reprocessing technologies came from the need for obtaining pure fissile material for nuclear weapon production. The most prominent among these is the PUREX (plutonium and uranium extraction) process, still used worldwide to reprocess commercial light water reactor fuels at a few thousand tons per year scale. The fuels dissolved in nitric acid are treated with a tributyl phosphate based solvent, the extracted uranium and plutonium are further purified, and the raffinate is vitrified for a safe final disposal. Plutonium is partly recycled as mixed oxide fuel. Since the beginning of this century, a new generation of nuclear reactors is being developed in the framework of the so-called generation IV initiative. For compliance with the sustainability goals defined for the innovative reactor systems, mainly waste minimization through recycling of all actinides, and for the achievement of these goals, the corresponding fuel cycles will play a central role. The new concept of a grouped actinide separation can be derived from aqueous or pyrochemical partitioning processes. For the aqueous schemes, a direct link to PUREX is obvious, namely, with coextraction of Np. The extraction of the remaining actinides can be achieved by using specially designed solvents based on phosphine oxide or diamide molecules. A major focus is on the very challenging separation of lanthanides from the trivalent actinides. The process implementation, especially for the less developed pyrometallurgy, requires a good understanding of the extraction mechanisms. Pyro-reprocessing, where all actinides are recycled, is based on metallic fuels; they are dissolved in molten salts at around 500–900 °C and actinides are selectively recovered, either by electrorefining or by extraction into a liquid metal phase. The fuels of new generation reactors will, at least in the beginning, most likely be oxides. Thus, for pyroprocesses a head-end reduction step for oxide into metals fuels is needed. A very specific reprocessing technology, the so-called direct use of pressurized water reactor spent fuel in CANDU process, is being developed in Korea. Here, used pressurized water reactor fuel is recycled to CANDU (CANada Deuterium Uranium) reactors after a dry treatment where volatile fission products are removed.

J.-P. Glatz

2012-01-01T23:59:59.000Z

228

Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

229

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

SciTech Connect

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.

Vinson, D.

2010-07-11T23:59:59.000Z

230

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

SciTech Connect

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.

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

1990-04-01T23:59:59.000Z

231

Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques -- Preliminary Modeling Results Emphasizing Integration among Techniques  

E-Print Network (OSTI)

Based Spent Fuel Analysis Safeguards Strategy Assessment,”on International Safeguards: Addressing VerificationNuclear Fuel,” European Safeguards R&D Association, Vilnius,

Tobin, S. J.

2010-01-01T23:59:59.000Z

232

Peer review at the University of Tromsø : a study of time spent on reviewing and researchers’ opinions on peer review .  

E-Print Network (OSTI)

??This thesis examines peer review at the University of Tromsø (UiT). It focuses on how much time is spent on reviewing at this particular institution,… (more)

Refsdal, Maria

2010-01-01T23:59:59.000Z

233

ITP Forest Products: Energy and Environmental Profile of the...  

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

Energy Use by Fuel for Paper Manufacturing Purchased Electricity 7% Fuel Oil 5% Natural Gas 18% Self Generated & Renewable Energy (Spent Pulping Liquors, Bark, Wood, Hydropower)...

234

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Aspects of Wet Storage of 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 nuclear fuel (SNF) generated by weapons and research programs and recovered through nonproliferation programs. The SNF consists of irradiated reactor fuel and cut up assemblies containing uranium, thorium and/or plutonium. The Department stores 34 metric tons of heavy metal SNF primarily

235

U.S. Spent Nuclear Fuel Data as of December 31, 1998  

Gasoline and Diesel Fuel Update (EIA)

Spent Nuclear Fuel Data, Detailed United States as of December 31, 1998 Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive Waste Management (OCRWM). The spent nuclear fuel (SNF) data includes detailed characteristics of SNF generated by commercial U.S. nuclear power plants. From 1983 through 1995 this data was collected annually. Since 1996 this data has been collected every three years. The latest available detailed data covers all SNF discharged from commercial reactors before December 31, 1998, and is maintained in a database by the EIA. Summary data tables from this database may be found as indicated below. The detailed data are available on request from Jim Finucane who can be reached at 202-287-1966 or at

236

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Basis for Identification of Disposal Options for R and D for Spent 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 barrier, and the waste. Salt, clay/shale, and granitic rocks represent a reasonable cross-section of behavior. Granitic rocks are also the primary basement rock to consider for deep borehole disposal. UFD is developing generic system analysis capability and general experimental data related to mined geologic disposal in the three

237

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

238

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies  

SciTech Connect

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01T23:59:59.000Z

239

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA  

SciTech Connect

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

2009-07-01T23:59:59.000Z

240

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Spent fuel and radioactive waste inventories, projections, and characteristics. Revision 1  

SciTech Connect

Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy (DOE) radioactive wastes were compiled through December 31, 1984, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the expected defense-related and private industrial and institutional activities and the latest DOE/Energy Information Administration (EIA) projections of US commercial nuclear power growth. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions.

Not Available

1985-12-01T23:59:59.000Z

242

Spent Fuel and High-Level Waste Requirements (Maine) | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Spent Fuel and High-Level Waste Requirements (Maine) Spent Fuel and High-Level Waste Requirements (Maine) Spent Fuel and High-Level Waste Requirements (Maine) < Back Eligibility Agricultural Commercial Construction Fed. Government Fuel Distributor General Public/Consumer Industrial Installer/Contractor Institutional Investor-Owned Utility Local Government Low-Income Residential Multi-Family Residential Municipal/Public Utility Nonprofit Residential Retail Supplier Rural Electric Cooperative Schools State/Provincial Govt Systems Integrator Transportation Tribal Government Utility Program Info State Maine Program Type Safety and Operational Guidelines Provider Public Utilities Commission All proposed nuclear power generation facilities must be certified by the Public Utilities Commission under this statute prior to construction and

243

Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety  

SciTech Connect

Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

DeHart, M.D.

1999-08-01T23:59:59.000Z

244

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01T23:59:59.000Z

245

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

SciTech Connect

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.

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.

2006-10-01T23:59:59.000Z

246

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

NONE

1994-03-25T23:59:59.000Z

247

An overview of measurement methods for special nuclear material in spent nuclear fuel  

SciTech Connect

Summary results from a survey of nondestructive assay measurement methods applicable to the measurement of the special nuclear material content of spent nuclear fuel are described. The role of nuclear materials measurements in the domestic and international safeguarding of spent nuclear fuel in the United States' federal waste management system has yet to be determined. An understanding of the characteristics and capabilities of the potentially applicable measurement systems should provide valuable information to the developers of the safeguards approaches for the monitored retrievable storage and final disposal systems. The discussion focuses on the general characteristics of the identified direct and indirect measurement methods. 3 refs., 1 tab.

Moran, B.W.; Reich, W.J.

1989-07-01T23:59:59.000Z

248

Retention of neptunium in uranyl alteration phases formed during spent fuel corrosion  

SciTech Connect

Uranyl oxide hydrate phases are known to form during contact of oxide spent nuclear fuel with water under oxidizing conditions; however, less is known about the fate of fission and neutron capture products during this alteration. We describe, the first time, evidence that neptunium can become incorporated into the uranyl secondary phase, dehydrated schoepite (UO{sub 3}{lg_bullet}0.8H{sub 2}O). Based on the long-term durability of natural schoepite, the retention of neptunium in this alteration phase may be significant during spent fuel corrosion in an unsaturated geologic repository.

Buck, E.C.; Finch, R.J.; Finn, P.A.; Bates, J.K.

1997-10-01T23:59:59.000Z

249

COBRA-SFS predictions of single assembly spent fuel heat transfer data  

SciTech Connect

The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments.

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

1986-04-01T23:59:59.000Z

250

A plasma signature of human mitochondrial disease revealed through metabolic profiling of spent media  

E-Print Network (OSTI)

A plasma signature of human mitochondrial disease revealed through metabolic profiling of spent culture with analysis of patient plasma to identify disease biomarkers. First, we applied global metabolic-defined metabolites in human plasma to discover that creatine is reproducibly elevated in two independent cohorts

Mootha, Vamsi K.

251

A numerical investigation of natural convection heat transfer within horizontal spent-fuel assemblies  

SciTech Connect

A numerical investigation of natural convection heat transfer is carried out for a single, horizontal, spent-fuel assembly in an environment typical of spent-fuel transportation systems as well as some dry storage/disposal scenarios. The objective is to predict computationally the convective heat transfer trends for horizontal spent fuel and to compare the results to data taken in a supporting experimental effort. The predicted data consist of thermal and flow fields throughout the assembly for a wide range of Rayleigh number, as well as numerically obtained Nusselt-number data that are correlated as a function of Rayleigh number. Both laminar and turbulent approaches are examined for a Boussinesq fluid with Pr = 0.7. The data predict the existence of a conduction-dominated regime, a transition regime, and a convection regime. Compared with the laminar approach, a significant improvement in the predicted Nusselt number is obtained for large Rayleigh numbers when a turbulence model is employed. This lends additional support to the experimental evidence that a transition to turbulent flow occurs for Rayleigh numbers greater than 10{sup 7}. Overall, the numerically predicted heat transfer trends compare well with previously obtained experimental data, and the computed assembly Nusselt numbers generally reside within the range of experimental uncertainty. The predicted thermal and flow fields further provide a numerical flow visualization capability that enhances the understanding of natural convection in horizontal spent fuel and allows improved physical interpretation of the experimental data.

Canaan, R.E. [Lawrence Livermore National Lab., CA (United States); Klein, D.E. [Univ. of Texas System, Austin, TX (United States)

1998-08-01T23:59:59.000Z

252

Radio-toxicity of spent fuel of the advanced heavy water reactor  

Science Journals Connector (OSTI)

......Radio-toxicity of spent fuel of the advanced heavy water reactor S. Anand * K. D. S...Mumbai 400085, India The Advanced Heavy Water Reactor (AHWR) is a new power...PHWR. INTRODUCTION The Advanced Heavy Water Reactor (AHWR)(1, 2), currently......

S. Anand; K. D. S. Singh; V. K. Sharma

2010-01-01T23:59:59.000Z

253

Author's personal copy Cost analysis of the US spent nuclear fuel reprocessing facility  

E-Print Network (OSTI)

of this, a 1987 amendment to the US Nuclear Waste Policy Act mandates the Secretary of Energy to report on a site for a second repository by 2010 (Nuclear Waste Policy Amendments Act, 1987). HoweverAuthor's personal copy Cost analysis of the US spent nuclear fuel reprocessing facility E

Deinert, Mark

254

Review of the natural circulation effect in the Vermont Yankee spent-fuel pool  

SciTech Connect

A 7429-node, three-dimensional computer model of the Vermont Yankee spent-fuel pool was set up and run using the porous media model of the TEMPEST computer code. The results of this analysis show that natural circulation is sufficient to ensure adequate cooling, regardless of the loading pattern used or the orientation of the cooling system discharge nozzle.

Wheeler, C.L.

1988-01-01T23:59:59.000Z

255

Dosimetry Modeling for Predicting Radiolytic Production at the Spent Fuel - Water Interface  

SciTech Connect

Modeling of the alpha, beta, and gamma dose from spent fuel as a function of particle size and fuel to water ratio was examined. These doses will be combined with modeling of G values and interactions to determine the concentration of various species formed at the fuel water interface and their affect on dissolution rates.

Miller, William H.; Kline, Amanda J.; Hanson, Brady D.

2006-04-30T23:59:59.000Z

256

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

SciTech Connect

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.

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

1992-04-01T23:59:59.000Z

257

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

SciTech Connect

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.

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

1992-04-01T23:59:59.000Z

258

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

SciTech Connect

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)

Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine)] [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine); Schmieman, Eric [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)] [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

259

Differential die-away technique for determination of the fissile contents in spent fuel assembly  

SciTech Connect

Monte Carlo simulations were performed for the differential die-away (DDA) technique to quantify its capability to measure the fissile contents in spent fuel assemblies of 64 different cases in terms of initial enrichment, burnup, and cooling time. The DDA count rate varies according to the contents of fissile isotopes such as {sup 235}U, {sup 239}Pu, and {sup 241}Pu contained in the spent fuel assembly. The effective {sup 239}Pu concept was introduced to quantify the total fissile mass of spent fuel by weighting the relative signal contributions of {sup 235}U and {sup 241}Pu compared to that of {sup 239}Pu. The Monte Carlo simulation results show that the count rate of the DDA instrument for a spent fuel assembly of 4% initial enrichment, 45 GWD/MTU burnup, and 5 year cooling time is {approx} 9.8 x 10{sup 4} counts per second (c/s) with the 100-Hz repeated interrogation pattern of 0 to 10 {micro}s interrogation, 0.2 ms to 1 ms counting time, and 1 x 10{sup 9} n/s neutron source. The {sup 244}Cm neutron background count rate for this counting time scheme is {approx} 1 x 10{sup 4} c/s, and thus the signal to background ratio is {approx}10.

Lee, Tachoon [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Nartyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

260

Radio-toxicity of spent fuel of the advanced heavy water reactor  

Science Journals Connector (OSTI)

......during their short/long term-storage is investigated in...radio-toxicity of radioactive waste is widely regarded...exchangers of the spent fuel storage bay. The decay power...VVER type reactors at long-term storage. Radiat. Prot. Dosim......

S. Anand; K. D. S. Singh; V. K. Sharma

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Home visiting yields as much as $5.70 for every dollar spent! 4  

E-Print Network (OSTI)

Home visiting yields as much as $5.70 for every dollar spent! 4 Home Why Florida Needs Home and specialized out-of-home care. Home visiting is an evidence-based strategy that can prevent these types the services. · Home visiting could improve Florida's low rank on child well-being. · Home visiting will save

Weston, Ken

262

Impact of PWR spent fuel variations on TRU-fueled VHTRS  

E-Print Network (OSTI)

of legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. Chapter III discusses the details of the study. The concluding study was accomplished through the use of ORNL SCALE5.1 code system. The detailed whole-core 3-D models were...

Alajo, Ayodeji Babatunde

2009-05-15T23:59:59.000Z

263

Assessment of the Fingerprinting Method for Spent Fuel Verification in MACSTOR KN-400 CANDU Dry Storage  

E-Print Network (OSTI)

The Korea Hydro and Nuclear Power has built a new modular type of dry storage facility, known as MACSTOR KN-400 at Wolsong reactor site. The building has the capacity to store up to 24000 CANDU spent fuel bundles in a 4 rows by 10 columns...

Gowthahalli Chandregowda, Nandan

2012-10-19T23:59:59.000Z

264

Hydrogen Storage by Boron?Nitrogen Heterocycles: A Simple Route for Spent Fuel Regeneration  

Science Journals Connector (OSTI)

We describe a new hydrogen storage platform based on well-defined BN heterocyle materials. Specifically, we demonstrate that regeneration of the spent fuel back to the charged fuel can be accomplished using molecular H2 and H?/H+ sources. Crystallographic ...

Patrick G. Campbell; Lev N. Zakharov; Daniel J. Grant; David A. Dixon; Shih-Yuan Liu

2010-02-19T23:59:59.000Z

265

Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study  

SciTech Connect

Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU.

McKinnon, Mikal A. (BATTELLE (PACIFIC NW LAB)); Cunningham, Mitchel E. (BATTELLE (PACIFIC NW LAB))

2003-09-09T23:59:59.000Z

266

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

SciTech Connect

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)

Parker, Frank L. [Vanderbilt University (United States)

2012-07-01T23:59:59.000Z

267

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

SciTech Connect

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).

Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

2008-03-01T23:59:59.000Z

268

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

SciTech Connect

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).

J. W. Sterbentz; D. L. Chichester

2011-09-01T23:59:59.000Z

269

Head-end process for the reprocessing of HTGR spent fuel  

SciTech Connect

The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

Chen, J.; Wen, M. [Institute of Nuclear and New Energy Technology, Tsinghua University, Bejing 10084 (China)

2013-07-01T23:59:59.000Z

270

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel  

SciTech Connect

Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron X-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO{sub 2}{sup 2+}) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of {approx} 50 micrometers. We find evidence of a thin ({approx} 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO{sub 2}{sup +}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. (authors)

Fortner, Jeffrey A.; Kropf, A. Jeremy; Jerden, James L.; Cunnane, James C. [Chemical Engineering, Argonne National Laboratory, CMT/205, 9700 S. Cass Avenue, Argonne, IL, 60439 (United States)

2007-07-01T23:59:59.000Z

271

Accelerated Dissolution Process of the Spent Fuel (UO{sub 2}) under Repository Conditions  

SciTech Connect

Nowadays, nuclear energy is one of the options for developed countries in order to maintain the demand of electric energy. One of the key problems associated with kind of energy generation is the residual waste formed after a fuel cycle (spent nuclear fuel). The thermal treatment received in the reactor and there composition renders these materials very difficult to characterize and thus exhaustive studies are required to obtain knowledge that will help to build a complete, reliable and very safety underground facility. In this way, the option known as the Deep Geological Repository (DGR) is under development by each country taking part in the nuclear energy industry. The unique pathway for the migration to the biosphere of the radionuclide, actinide and lanthanides content in the spent fuel pellet (UO{sub 2}) after the closing of the deep geological repository is by a water transport phenomena. It is a fundamental question to know how much time they will spend on their trip and the first step is the rate of liberation of these radionuclides from the spent fuel pellet. In this way the matrix dissolution rate of the spent fuel pellet, which is not dependent on the specific surface area after normalization by the initial value is a key parameter to begin the performance assessment for any deep geological repository. The specific surface value is, following the Matrix Alteration Model (MAM) sensitivity analysis, one of the most important parameters controlling the radionuclides liberation. In this way, several measurements were carried out to obtain values in different conditions for different sieves of UO{sub 2} powder treated as fresh fuel. First of all, the specific surface area was measured with a multi-point isothermal procedure with N{sub 2} and Kr for both. The values obtained are presented in order to obtain a general law for the rate of evolution with the particle size. These data are part of a bigger project about the complete description of the spent fuel analogues, which are very useful for obtaining new dissolution rates for spent nuclear fuel under repository simulated conditions. (authors)

Iglesias, Eduardo; Quinones, Javier; Rodriguez, Nieves [Energy, CIEMAT, Avda. Complutense 22, Madrid, 28040 (Spain)

2008-07-01T23:59:59.000Z

272

Environmental Statements, Availability, Etc., Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8679 8679 Thursday June 1, 1995 Part III Department of Energy Environmental Statements, Availability, Etc.; Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs: Notice 28680 Federal Register / Vol. 60, No. 105 / Thursday, June 1, 1995 / Notices DEPARTMENT OF ENERGY Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: The Department of Energy has issued a Record of Decision on Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs. The Record of Decision includes a Department-wide decision to

273

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel  

NLE Websites -- All DOE Office Websites (Extended Search)

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Vitaly T. Gotovchikov a , Victor A. Seredenko a , Valentin V. Shatalov a , Vladimir N. Kaplenkov a , Alexander S. Shulgin a , Vladimir K. Saranchin a , Michail A. Borik a∗ , Charles W. Forsberg b , All-Russian Research Institute of Chemical Technology (ARRICT) 33, Kashirskoe ave., Moscow, Russia, 115409, E-mail: chem.conv@ru.net Oak Ridge National Laboratory (ORNL) Bethel Wall Road, P.O. Box 2008, MS-6165, Oak Ridge, TN, USA, 37831 Abstract - Induction cold crucible melters (ICCM) have the potential to be a very-low-cost high-throughput method for the production of DUO 2 for SNF casks. The proposed work would develop these melters for this specific application. If a

274

SPENT NUCLEAR FUEL STORAGE BASIN WATER CHEMISTRY: ELECTROCHEMICAL EVALUATION OF ALUMINUM CORROSION  

SciTech Connect

The factors affecting the optimal water chemistry of the Savannah River Site spent fuel storage basin must be determines in order to optimize facility efficiency, minimize fuel corrosion, and reduce overall environmental impact from long term spent nuclear fuel storage at the Savannah River Site. The Savannah River National Laboratory is using statistically designed experiments to study the effects of NO{sub 3}{sup -}, SO{sub 4}{sup 2-}, and Cl{sup -} concentrations on alloys commonly used not only as fuel cladding, but also as rack construction materials The results of cyclic polarization pitting and corrosion experiments on samples of Al 6061 and 1100 alloys will be used to construct a predictive model of the basin corrosion and its dependence on the species in the basin. The basin chemistry model and corrosion will be discussed in terms of optimized water chemistry envelope and minimization of cladding corrosion.

Hathcock, D

2007-10-30T23:59:59.000Z

275

Proposed subcritical measurements for fresh and spent highly enriched plate type fuel assemblies  

SciTech Connect

A collaborative experimental research program has been established between industry and university partners to evaluate the subcritical behavior of fresh and spent highly enriched fuel assemblies at the University of Missouri Research Reactor (MURR). This proposed program will involve a series of subcritical measurements using the Oak Ridge National Laboratory (ORNL) developed {sup 252}Cf source-driven noise technique. Measurements evaluating the subcritical behavior of simple arrays of fresh MURR assemblies will be performed for evaluating the spectral effects of materials typically found in shipping casks such as lead, steel, aluminum, and boron. Also, measurements will be performed on spent assemblies to characterize physics parameters which may be useful in determining the subcritical behavior of fuels for reactivity credit of actinide burnup and fission product poisoning.

Zino, J.F.; Williamson, T.G. [Westinghouse Savannah River Company, Aiken, SC (United States); Mihalczo, J.T. [Oak Ridge National Lab., TN (United States)] [and others

1997-09-01T23:59:59.000Z

276

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

SciTech Connect

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.

Khraisha, Y.H.

2000-05-01T23:59:59.000Z

277

Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel  

SciTech Connect

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

L. Angers

2001-01-31T23:59:59.000Z

278

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

SciTech Connect

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.

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

2002-01-30T23:59:59.000Z

279

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

SciTech Connect

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.

Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

1992-11-01T23:59:59.000Z

280

Probabilistic assessment of spent fuel shipping cask response to severe transportation accident conditions. Report summary  

SciTech Connect

The licensing of commercial nuclear spent shipping casks in the United States is regulated by 10CFR71. In order to be licensed, casks must be designed not to fail under hypothetical test conditions specified in Appendix B of this regulation. Questions have been raised about the suitability of these tests in simulating actual transportation accident conditions. Our study addresses the adequacy of current regulations by comparing real-world accident conditions with regulatory test specifications using more complete accident statistics and more sophisticated structural analyses than have been used in studies to date. Our objective is to evaluate the protection provided by current regulations against severe accident conditions for commercial spent nuclear fuel casks that are transported by truck or rail. The complete spectrum of truck and rail accidents will be reviewed in order to determine the frequency (or infrequency) of cask failures during transportation accidents. 3 references, 1 figure.

Fischer, L.E.; Kimura, C.Y.; Witte, M.C.

1985-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

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

SciTech Connect

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.

Ahn, T.M.

1996-11-01T23:59:59.000Z

282

On the Development of a Distillation Process for the Electrometallurgical Treatment of Irradiated Spent Nuclear Fuel  

SciTech Connect

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

B.R. Westphal; K.C. Marsden; J.C. Price; D.V. Laug

2008-04-01T23:59:59.000Z

283

Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints  

SciTech Connect

The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

1989-08-01T23:59:59.000Z

284

Annual Radiological Environmental Monitoring Program Report for the Fort St. Vrain Independent Spent Fuel Storage Installation  

SciTech Connect

This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Fort St. Vrain Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the predominant radiation exposure pathway, direct and scattered radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

Hall, Gregory Graham; Newkirk, Jay Ronald; Borst, Frederick Jon

2002-02-01T23:59:59.000Z

285

Annual Radiological Environmental Monitoring Program Report for the Fort St. Vrain Independent Spent Fuel Storage Installation  

SciTech Connect

This report presents the results of the 2002 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Fort St. Vrain Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the predominant radiation exposure pathway, direct and scattered radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

Jay R. Newkirk; Frederick J. Borst; Gregory G. Hall

2003-02-01T23:59:59.000Z

286

A comparison of spent fuel assembly control instruments: The Cadarache PYTHON and the Los Alamos Fork  

SciTech Connect

Devices to monitor spent fuel assemblies while stored under water with nondestructive assay methods, have been developed in France and in the United States. Both devices are designed to verify operator's declared values of exposures and cooling-time but the applications and thus the designs of the systems differ. A study, whose results are presented in this paper, has been conducted to compare the features and the performances of the two instruments. 4 refs., 9 figs.

Bignan, G.; Capsie, J.; Romeyer-Dherbey, J. (CEA Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires); Rinard, P. (Los Alamos National Lab., NM (United States))

1991-01-01T23:59:59.000Z

287

K Basin spent nuclear fuel hot conditioning system functions {ampersand} requirements  

SciTech Connect

The purpose of this F{ampersand}R document is to establish the functional requirements baseline for the Spent Nuclear Fuel Hot Conditioning System (HCS) subproject. This F{ampersand}R documents the: -mission of the HCS, -evolution of the technical baseline leading to the HCS, -functions that must be performed to accomplish the HCS mission, -requirements basis allocated to the HCS mission and functions, -identification and definition of interfaces between the HCS and other SNF subprojects.

Miska, C.R., Westinghouse Hanford

1996-07-08T23:59:59.000Z

288

Evaluation of potential food applications of dried distillers spent grain (DSG). Final research report  

SciTech Connect

Results from experimental test bakes indicate that dried distillers spent grain (DSG) can be used to replace up to 15% of the flour for the production of an acceptable variety bread, provided that the DSG is processed under optimum conditions for a satisfactory flavor and color development. The raw materials used for the mash bill may also affect the taste of the finished product. (MHR)

Not Available

1981-02-06T23:59:59.000Z

289

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

SciTech Connect

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.

Linge, I. I.; Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

2012-12-15T23:59:59.000Z

290

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

SciTech Connect

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.

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

1980-10-11T23:59:59.000Z

291

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

SciTech Connect

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.

Cardoni, Jeffrey

2010-11-01T23:59:59.000Z

292

The Effect of Acid Additives on Carbonate Rock Wettability and Spent Acid Recovery in Low Permeability Gas Carbonates  

E-Print Network (OSTI)

Spent acid retention in the near-wellbore region causes reduction of relative permeability to gas and eventually curtailed gas production. In low-permeability gas carbonate reservoirs, capillary forces are the key parameters that affect the trapping...

Saneifar, Mehrnoosh

2012-10-19T23:59:59.000Z

293

UTILIZATION OF SPENT RADIOISOTOPE THERMOELECTRIC GENERATORS AND INSTALLATION OF SOLAR CELL TECHNOLOGY AS POWER SOURCE FOR RUSSIAN LIGHTHOUSES - FINAL REPORT  

Science Journals Connector (OSTI)

The Northern Fleets hydrographical department has with support from Norway worked on the utilization of spent strontium-containing RTGs used as power sources at lighthouses situated at the Kola Peninsula.

PER-EINAR FISKEBECK

2006-01-01T23:59:59.000Z

294

Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs  

SciTech Connect

This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

1987-09-01T23:59:59.000Z

295

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

SciTech Connect

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.

Ilas, Germina [ORNL; Gauld, Ian C [ORNL

2011-01-01T23:59:59.000Z

296

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

DOE Patents (OSTI)

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.

Fish, Richard H. (Berkeley, CA)

1987-01-01T23:59:59.000Z

297

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

DOE Patents (OSTI)

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.

Fish, R.H.

1987-04-21T23:59:59.000Z

298

Development of a techno-economic model to optimization DOE spent nuclear fuel disposition  

SciTech Connect

The purpose of the National Spent Nuclear Fuel (NSNF) Program conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL) is to evaluate what to do with the spent nuclear fuel (SNF) in the Department of Energy (DOE) complex. Final disposition of the SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on the fuel type and the current locations of the fuel. One of the first steps associated with selecting one or more sites for treating the SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the electrometallurgical treatment process for fuels at several locations. The set of questions addresses all issues associated with the design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs will be applied to determine the life-cycle cost of each option. This technique can also be applied to other treatment techniques for treating spent nuclear fuel.

Ramer, R.J.; Plum, M.M.; Adams, J.P.; Dahl, C.A.

1997-11-01T23:59:59.000Z

299

Spent fuel sabotage test program, characterization of aerosol dispersal : technical review and analysis supplement.  

SciTech Connect

This project seeks to provide vital data required to assess the consequences of a terrorist attack on a spent fuel transportation cask. One such attack scenario involves the use of conical shaped charges (CSC), which are capable of damaging a spent fuel transportation cask. In the event of such an attack, the amount of radioactivity that may be released as respirable aerosols is not known with great certainty. Research to date has focused on measuring the aerosol release from single short surrogate fuel rodlets subjected to attack by a small CSC device in various aerosol chamber designs. The last series of three experiments tested surrogate fuel rodlets made with depleted uranium oxide ceramic pellets in a specially designed double chamber aerosol containment apparatus. This robust testing apparatus was designed to prevent any radioactive release and allow high level radioactive waste disposal of the entire apparatus following testing of actual spent fuel rodlets as proposed. DOE and Sandia reviews of the project to date identified a number of issues. The purpose of this supplemental report is to address and document the DOE review comments and to resolve the issues identified in the Sandia technical review.

Durbin, Samuel G.; Lindgren, Eric Richard

2009-07-01T23:59:59.000Z

300

Probable leaching mechanisms for UO/sub 2/ and spent fuel  

SciTech Connect

The oxidation and dissolution mechanisms for UO/sub 2/ and spent fuel will be quite similar based on this preliminary work with electrochemical leaching of UO/sub 2/ and spent fuel. In solutions containing oxygen or other oxidizing species, the UO/sub 2/ surface will be rapidly oxidized and dissolved following the transformation of uranium from U(IV) to U(VI). The hydrolysis of dissolved uranyl ions forms solid UO/sub 3/ hydrates or related complex compounds deposited onto the UO/sub 2/ surface, or other surfaces, as thin or thick coatings. Depending on the pH, temperature, and time, the various kinds of porosity and the mechanical properties of the hydrate coatings will control the dissolution rate. The effects of radiation, in terms of generation of H/sub 2/O/sub 2/, will enhance the dissolution kinetics. Electrochemical methods may be useful for determining the surface conditions, dissolution rate, and accelerated dissolution behavior for NO/sub 2/ and spent fuel. Electrochemial methods can rapidly generate much information in terms of dissolution rate and surface film properties - such as thickness, porosity, and oxidation state - in-situ during the leaching process.

Wang, R.; Katayama, Y.B.

1980-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Integrated data base for 1986: spent fuel and radioactive waste inventories, projections, and characteristics. Revision 2  

SciTech Connect

The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US Department of Energy (DOE) radioactive wastes through December 31, 1985, based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. Current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the expected defense-related and private industrial and institutional activities and the latest DOE/Energy Information Administration (EIA) projections of US commercial nuclear power growth. The materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or calculated isotopic compositions.

Not Available

1986-09-01T23:59:59.000Z

302

Integrated Data Base for 1989: Spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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, 1988. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. 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 defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, commercial reactor and fuel cycle facility decommissioning waste, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, 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, highly radioactive materials that may require geologic disposal. 45 figs., 119 tabs.

Not Available

1989-11-01T23:59:59.000Z

303

Integrated data base for 1990: US spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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, 1989. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. 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 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 2020, 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. 22 refs., 48 figs., 109 tabs.

Not Available

1990-10-01T23:59:59.000Z

304

Integrated data base for 1988: Spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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, 1987. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. 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 defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reportd for miscellaneous, highly radioactive materials that may require geologic disposal. 89 refs., 46 figs., 104 tabs.

Not Available

1988-09-01T23:59:59.000Z

305

Integrated data base for 1987: Spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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, 1986. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. Current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. 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 defense-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, 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, highly radioactive materials that may require geologic disposal. 82 refs., 57 figs., 121 tabs.

Not Available

1987-09-01T23:59:59.000Z

306

Effects of an oxidizing atmosphere in a spent fuel packaging facility  

SciTech Connect

Sufficient oxidation of spent fuel can cause a cladding breach to propagate, resulting in dispersion of fuel particulates and gaseous radionuclides. The literature for spent fuel oxidation in storage and disposal programs was reviewed to evaluate the effect of an oxidizing atmosphere in a preclosure packaging facility on (1) physical condition of the fuel and (2) operations in the facility. Effects such as cladding breach propagation, cladding oxidation, rod dilation, fuel dispersal, {sup 14}C and {sup 85}Kr release, and crud release were evaluated. The impact of these effects, due to oxidation, upon a spent fuel handling facility is generally predicted to be less than the impact of similar effects due to fuel rod breached during handling in an inert-atmosphere facility. Preliminary temperature limits of 240{degree}C and 227{degree}C for a 2-week or 4-week handling period and 175{degree}C for 2-year lag storage would prevent breach propagation and fuel dispersal. Additional data that are needed to support the assumptions in this analysis or complete the database were identified.

Einziger, R.E.

1991-09-01T23:59:59.000Z

307

Dry halide method for separating the components of spent nuclear fuels  

DOE Patents (OSTI)

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.

Christian, Jerry Dale (Idaho Falls, ID); Thomas, Thomas Russell (Rigby, ID); Kessinger, Glen F. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

308

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

SciTech Connect

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.

M.R. Scott

2005-06-01T23:59:59.000Z

309

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SUPPLEMENT ANALYSIS FOR THE FOREIGN 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. Background .............................................................................................................................................. 1 3. The Proposed Action ...............................................................................................................................

310

Initial measurements of BN-350 spent fuel in dry storage casks using the dual slab verification detonator  

SciTech Connect

The Dual Slab Verification Detector (DSVD) has been developed, built, and characterized by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of 3He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. By performing DSVD measurements at several different locations around the outer surface of the DUC, a signature 'fingerprint' can be established for each DUC based on the neutron flux emanating from inside the dry storage cask. The neutron fingerprint for each individual DUC will be dependent upon the spatial distribution of nuclear material within the cask, thus making it sensitive to the removal of a certain amount of material from the cask. An initial set of DSVD measurements have been performed on the first set of dry storage casks that have been loaded with canisters of spent fuel and moved onto the dry storage pad to both establish an initial fingerprint for these casks as well as to quantify systematic uncertainties associated with these measurements. The results from these measurements will be presented and compared with the expected results that were determined based on MCNPX simulations of the dry storage facility. The ability to safeguard spent nuclear fuel is strongly dependent on the technical capabilities of establishing and maintaining continuity of knowledge (COK) of the spent fuel as it is released from the reactor core and either reprocessed or packaged and stored at a storage facility. While the maintenance of COK is often done using continuous containment and surveillance (C/S) on the spent fuel, it is important that the measurement capabilities exist to re-establish the COK in the event of a significant gap in the continuous CIS by performing measurements that independently confirm the presence and content of Plutonium (Pu) in the spent fuel. The types of non-destructive assay (NDA) measurements that can be performed on the spent fuel are strongly dependent on the type of spent fuel that is being safeguarded as well as the location in which the spent fuel is being stored. The BN-350 Spent Fuel Disposition Project was initiated to improve the safeguards and security of the spent nuclear fuel from the BN-350 fast-breeder reactor and was developed cooperatively to meet the requirements of the International Atomic Energy Agency (IAEA) as well as the terms of the 1993 CTR and MPC&A Implementing Agreements. The unique characteristics of fuel from the BN-350 fast-breeder reactor have allowed for the development of an integrated safeguards measurement program to inventory, monitor, and if necessary, re-verify Pu content of the spent fuel throughout the lifetime of the project. This approach includes the development of a safeguards measurement program to establish and maintain the COK on the spent fuel during the repackaging and eventual relocation of the spent-fuel assemblies to a long-term storage site. As part of the safeguards measurement program, the Pu content of every spent-fuel assembly from the BN-350 reactor was directly measured and characterized while the spent-fuel assemblies were being stored in the spent-fuel pond at the BN-350 facility using the Spent Fuel Coincidence Counter (SFCC). Upon completion of the initial inventory of the Pu content of the individual spent-fuel assemblies, the assemblies were repackaged into welded steel canisters that were filled with inert argon gas and held either four or six individual spent-fuel assemblies depending on the type of assembly that was being packaged. This repackaging of the spent-fuel assemblies was performed in order to improve the stability of the spent-fuel assemblies for long-term storage and increase the proliferation resistance of the spent fuel. To maintain the capability of verifying the presence of the spent-fuel assemblies inside the welded steel canisters, measurements were performed on the canis

Santi, Peter Angelo [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Freeman, Corey R [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

311

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

SciTech Connect

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.

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

2012-06-04T23:59:59.000Z

312

Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions  

SciTech Connect

Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10{sup {minus}6} per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool.

Glumac, B; Ravnik, M. [Jozef Stefan Inst., Ljubljana (Slovenia); Logar, M. [Univ. of Maribor (Slovenia). Faculty of Electrical Engineering and Computer Sciences

1997-02-01T23:59:59.000Z

313

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL  

SciTech Connect

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

2003-07-01T23:59:59.000Z

314

Realization of the German Concept for Interim Storage of Spent Nuclear Fuel - Current Situation and Prospects  

SciTech Connect

The German government has determined a phase out of nuclear power. With respect to the management of spent fuel it was decided to terminate transports to reprocessing plants by 2005 and to set up interim storage facilities on power plant sites. This paper gives an overview of the German concept for spent fuel management focused on the new on-site interim storage concept and the applied interim storage facilities. Since the end of the year 1998, the utilities have applied for permission of on-site interim storage in 13 storage facilities and 5 storage areas; one application for the interim storage facility Stade was withdrawn due to the planned final shut down of Stade nuclear power plant in autumn 2003. In 2001 and 2002, 3 on-site storage areas and 2 on-site storage facilities for spent fuel were licensed by the Federal Office for Radiation Protection (BfS). A main task in 2002 and 2003 has been the examination of the safety and security of the planned interim storage facilities and the verification of the licensing prerequisites. In the aftermath of September 11, 2001, BfS has also examined the attack with a big passenger airplane. Up to now, these aircraft crash analyses have been performed for three on-site interim storage facilities; the fundamental results will be presented. It is the objective of BfS to conclude the licensing procedures for the applied on-site interim storage facilities in 2003. With an assumed construction period for the storage buildings of about two years, the on-site interim storage facilities could then be available in the year 2005.

Thomauske, B. R.

2003-02-25T23:59:59.000Z

315

Effects of spent fuel types on offsite consequences of hypothetical accidents  

SciTech Connect

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.

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

2000-02-18T23:59:59.000Z

316

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel  

SciTech Connect

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.

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

2009-09-01T23:59:59.000Z

317

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.  

SciTech Connect

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-02-17T23:59:59.000Z

318

Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)  

SciTech Connect

Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and angle of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.

Koji Shirai; Jyunichi Tani; Taku Arai; Masumi Watatu; Hirofumi Takeda; Toshiari Saegusa; Philip L. Winston

2008-10-01T23:59:59.000Z

319

Facts and issues of direct disposal of spent fuel; Revision 1  

SciTech Connect

This report reviews those facts and issues that affect the direct disposal of spent reactor fuels. It is intended as a resource document for those impacted by the current Department of Energy (DOE) guidance that calls for the cessation of fuel reprocessing. It is not intended as a study of the specific impacts (schedules and costs) to the Savannah River Site (SRS) alone. Commercial fuels, other low enriched fuels, highly enriched defense-production, research, and naval reactor fuels are included in this survey, except as prevented by rules on classification.

Parks, P.B.

1993-10-01T23:59:59.000Z

320

Experience With Damaged Spent Nuclear Fuel at U.S. DOE Facilities  

SciTech Connect

This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report. (authors)

Carlsen, Brett; Fillmore, Denzel; Woolstenhulme, Eric [Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415 (United States); McCormack, Roger L. [Fluor Hanford Site, Richland, Wash. (United States); Sindelar, Robert; Spieker, Timothy [Savannah River National Laboratory, Savannah River Site Aiken, SC 29808 (United States)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Non-destructive assay of spent nuclear fuel using passive neutron Albedo reactivity  

SciTech Connect

Passive Neutron Albedo Reactivity (PNAR) is one of fourteen techniques that has been researched and evaluated to form part of a comprehensive and integrated detection system for the non-destructive assay (NDA) of spent nuclear fuel. PNAR implemented with {sup 3}He tubes for neutron detection (PNAR-{sup 3}He) is the measurement of time correlated neutrons from a spent fuel assembly with and without a Cadmium (Cd) layer surrounding the assembly. PNAR utilizes the self-interrogation of the fuel via reflection of neutrons born in the fuel assembly back in to the fuel assembly. The neutrons originate primarily from spontaneous fission events within the fuel itself (Curium-244) but are amplified by multiplication. The presence and removal of the Cd provides two measurement conditions with different neutron energy spectra and therefore different interrogating neutron characteristics. Cd has a high cross-section of absorption for slow neutrons and therefore greatly reduces the low energy (thermal) neutron fluence rate returning. The ratios of the Singles, Doubles and Triples count rates obtained in each case are known as the Cd ratios, which are related to fissile content. A potential safeguards application for which PNAR-{sup 3}He is particularly suited is 'fingerprinting'. Fingerprinting could function as an alternative to plutonium (Pu) mass determination; providing confidence that material was not diverted during transport between sites. PNAR-{sup 3}He has six primary NDA signatures: Singles, Doubles and Triples count rates measured with two energy spectra at both shipping and receiving sites. This is to uniquely identify the fuel assembly, and confirm no changes have taken place during transport. Changes may indicate all attempt to divert material for example. Here, the physics of the PNAR-{sup 3}He concept will be explained, alongside a discussion on the development of a prototypical PNAR-{sup 3}He instrument using simulation. The capabilities and performance of the conceptual instrument will be summarized, in the context of (a) quantifying Pu mass in spent fuel assemblies and (b) detecting pin diversion (through a discrepancy between declared and measured properties of the fuel assembly) when the instrument is deployed. These quantitative capabilities are complementary to the 'fingerprinting' capability which is part of ensuring continuity of knowledge and custody of spent nuclear fuel.

Evans, L G [Los Alamos National Laboratory; Schear, M A [Los Alamos National Laboratory; Croft, S [Los Alamos National Laboratory; Tobin, S J [Los Alamos National Laboratory; Swinhoe, M T [Los Alamos National Laboratory; Menlove, H O [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

322

Spent Isopropanol Solution as Possible Liquid Fuel for Moving Bed Reactor in Chemical Looping Combustion  

Science Journals Connector (OSTI)

Spent Isopropanol Solution as Possible Liquid Fuel for Moving Bed Reactor in Chemical Looping Combustion ... The fuels, such as natural gas, coal, petroleum coke, and biomass combusted by CLC are frequently studied by various researchers(17, 26-31) and compared in the previous studies;(20, 33) however, only few studies on liquid fuel combustion are reported. ... Ishida, M.; Takeshita, K.; Susuki, K.; Ohba, T..Application of Fe2O3-Al2O3 composite particles as solid looping material of the chemical loop combustor Energy Fuels 2005, 19, 2514– 2518 ...

Ping-Chin Chiu; Young Ku; Hsuan-Chih Wu; Yu-Lin Kuo; Yao-Hsuan Tseng

2013-10-31T23:59:59.000Z

323

What are Spent Nuclear Fuel and High-Level Radioactive Waste ?  

SciTech Connect

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

DOE

2002-12-01T23:59:59.000Z

324

Considerations for increasing unit 1 spent fuel pool capacity at the Laguna Verde station  

SciTech Connect

To increase the spent fuel storage capacity at the Laguna Verde Station in a safe and economical manner and assure a continuous operation of the first Mexican Nuclear Plant, Comision Federal de Electricidad (CFE), the Nation's Utility, seeked alternatives considering the overall world situation, the safety and licensing aspects, as well as the economics and the extent of the nuclear program of Mexico. This paper describes the alternatives considered, their evaluation and how the decision taken by CFE in this field, provides the Laguna Verde Station with a maximum of 37 years storage capacity plus full core reserve.

Vera, A. (Comision Federal de Electricidad, Veracruz, Ver. (Mexico))

1992-01-01T23:59:59.000Z

325

Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel.  

SciTech Connect

Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to {approx} 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium.

Janney, D. E.; Keiser, D. D., Jr.; Engineering Technology

2003-09-01T23:59:59.000Z

326

Spent Nuclear Fuel Project technical baseline document. Fiscal year 1995: Volume 1, Baseline description  

SciTech Connect

This document is a revision to WHC-SD-SNF-SD-002, and is issued to support the individual projects that make up the Spent Nuclear Fuel Project in the lower-tier functions, requirements, interfaces, and technical baseline items. It presents results of engineering analyses since Sept. 1994. The mission of the SNFP on the Hanford site is to provide safety, economic, environmentally sound management of Hanford SNF in a manner that stages it to final disposition. This particularly involves K Basin fuel, although other SNF is involved also.

Womack, J.C. [Westinghouse Hanford Co., Richland, WA (United States); Cramond, R. [TRW (United States); Paedon, R.J. [SAIC (United States)] [and others

1995-03-13T23:59:59.000Z

327

Kinematics and thermodynamics across a propagating non-stoichiometric oxidation phase front in spent fuel grains  

SciTech Connect

Spent fuel contains mixtures, alloy and compound, but are dominated by U and O except for some UO{sub 2} fuels with burnable poisons (gadolinia in BWR rods), the other elements evolve during reactor operation from neutron reaction and fission + fission decay events. Due to decay, chemical composition and activity of spent fuel will continue to evolve after removal from reactors. During the time interval with significant radioactivity levels relevant for a geological repository, it is important to develop models for potential chemical responses in spent fuel and potential degradation of repository. One such potential impact is the oxidation of spent fuel, which results in initial phase change of UO{sub 2} lattice to U{sub 4}O{sub 9} and the next phase change is probably to U{sub 3}O{sub 8} although it has not been observed yet below 200C. The U{sub 4}O{sub 9} lattice is nonstoichiometric with a O/U weight ratio at 2.4. Preliminary indications are that the UO{sub 2} has a O/U of 2. 4 at the time just before it transforms into the U{sub 4}O{sub 9} phase. In the oxygen weight gain versus time response, a plateau appears as the O/U approaches 2.4. Part of this plateau is due to geometrical effects of a U{sub 4}O{sub 9} phase change front propagating into UO{sub 2} grain volumes; however, this may indicate a metastable phase change delay kinetics or a diffusional related delay time until the oxygen density can satisfy stoichiometry and energy conditions for phase changes. Experimental data show a front of U{sub 4}O{sub 9} lattice structure propagating into grains of the UO{sub 2} lattice. To describe this spatially inhomogenous oxidation phase transition, as well as the expected U{sub 3}O{sub 8} phase transition from the U{sub 4}O{sub 9} lattice, lattice models are developed and spatially discontinuous kinematic and energetic expressions are derived. 9 refs.

Stout, R.B.; Kansa, E.J.; Wijesinghe, A.M.

1993-09-01T23:59:59.000Z

328

Monte Carlo simulations of a differential die-away instrument for determination of fissile content in spent fuel assemblies  

SciTech Connect

The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability to measure the fissile content in spent fuel assemblies, For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighting the relative contribution to the signal of {sup 235}U and {sup 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

Lee, Taehoon [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

329

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

Energy.gov (U.S. Department of Energy (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.

330

Centrifugal microfluidic platform for radiochemistry: Potentialities for the chemical analysis of nuclear spent fuels  

Science Journals Connector (OSTI)

Abstract The use of a centrifugal microfluidic platform is for the first time reported as an alternative to classical chromatographic procedures for radiochemistry. The original design of the microfluidic platform has been thought to fasten and simplify the prototyping process with the use of a circular platform integrating four rectangular microchips made of thermoplastic. The microchips, dedicated to anion-exchange chromatographic separations, integrate a localized monolithic stationary phase as well as injection and collection reservoirs. The results presented here were obtained with a simplified simulated nuclear spent fuel sample composed of non-radioactive isotopes of Europium and Uranium, in proportion usually found for uranium oxide nuclear spent fuel. While keeping the analytical results consistent with the conventional procedure (extraction yield for Europium of ?97%), the use of the centrifugal microfluidic platform allowed to reduce the volume of liquid needed by a factor of ?250. Thanks to their unique “easy-to-use” features, centrifugal microfluidic platforms are potential successful candidates for the downscaling of chromatographic separation of radioactive samples (automation, multiplexing, easy integration in glove-boxes environment and low cost of maintenance).

Anthony Bruchet; Vélan Taniga; Stéphanie Descroix; Laurent Malaquin; Florence Goutelard; Clarisse Mariet

2013-01-01T23:59:59.000Z

331

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

SciTech Connect

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.

D.R. Jackson; G.R. Kiebel

1999-08-24T23:59:59.000Z

332

Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium  

Science Journals Connector (OSTI)

The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D–T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

Yousry Gohar

2001-01-01T23:59:59.000Z

333

Electrolytic Reduction of Spent Light Water Reactor Fuel Bench-Scale Experiment Results  

SciTech Connect

A series of experiments were performed to demonstrate the electrolytic reduction of spent light water reactor fuel at bench-scale in a hot cell at the Idaho National Laboratory Materials and Fuels Complex. The process involves the conversion of oxide fuel to metal by electrolytic means, which would then enable subsequent separation and recovery of actinides via existing electrometallurgical technologies, i.e., electrorefining. Four electrolytic reduction runs were performed at bench scale using ~500 ml of molten LiCl – 1 wt% Li2O electrolyte at 650 ºC. In each run, ~50 g of crushed spent oxide fuel was loaded into a permeable stainless steel basket and immersed into the electrolyte as the cathode. A spiral wound platinum wire was immersed into the electrolyte as the anode. When a controlled electric current was conducted through the anode and cathode, the oxide fuel was reduced to metal in the basket and oxygen gas was evolved at the anode. Salt samples were extracted before and after each electrolytic reduction run and analyzed for fuel and fission product constituents. The fuel baskets following each run were sectioned and the fuel was sampled, revealing an extent of uranium oxide reduction in excess of 98%.

Steven D. Herrmann

2007-04-01T23:59:59.000Z

334

Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect

Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed.

MASSEY,CHARLES D.; MESSICK,C.E.; MUSTIN,T.

1999-11-01T23:59:59.000Z

335

Development of a Techno-Economic Model to Optimize DOE Spent Nuclear Fuel Disposition  

SciTech Connect

The National Spent Nuclear Fuel (NSNF) Program is evaluating final disposition of spent nuclear fuel (SNE) in the Department of Energy (DOE) complex. Final disposition of SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment (EMT) and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on fuel type and location of the fuel. One of the first steps associated with selecting one or more sites for treating SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the EMT process for fuels at several locations. The set of questions addresses all issues associated with design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs can be applied to determine the life cycle cost of each option. This technique can also be applied to other treatment techniques for treating SNF.

Ramer, R. J.; Plum, M. M.; Adams, J. P.; Dahl, C. A.

1998-02-01T23:59:59.000Z

336

Investigations of Occidental Oil Shale, Inc. , retort 3E spent shales  

SciTech Connect

The mineralogy, trace element concentrations in the solids, and leaching characteristics of spent shales from retort 3E cores were studied. This is the first comprehensive study of in situ generated materials. Although there were unique aspects to the operations of retort 3E that are not representative of a commercial operation, the study of in situ generated materials has provided insights into the chemical, mineralogic, and solution reactions that occur during and subsequent to underground retorting. Characterization of the solid materials has verified that in situ processing occurs in a thermodynamic and kinetic regime not previously encountered for surface retorting. Carbonate decomposition and silication reactions that form high-temperature products, such as akermanite-gehlenite and diopside-augite solid solutions, have been identified by x-ray diffraction. Observations resulting from solids characterization are substantiated by leachate chemistry, which, although compatible with leachates generated from surface retorted materials, are different in extent (composition and concentration). Comparison of leachates generated from in situ materials with Multi-Media Environmental Goals/Minimum Acute Toxicity Effluent values indicates that the major and trace elements deserving further investigation are potassium, lithium, fluoride, vanadium, boron, molybdenum, nickel, and arsenic. Several trace elements, including fluoride, vanadium, boron, and arsenic, show increased mobilities from spent shales containing akermanite-gehlenite solid solutions as the major mineral phase. This report details the chemical, mineralogic, and solution behavior of major, minor, and trace elements in these minerals.

Peterson, E.J.; Henicksman, A.; Wagner, P.

1981-06-01T23:59:59.000Z

337

Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates  

SciTech Connect

This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

1996-05-01T23:59:59.000Z

338

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

SciTech Connect

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.

Durbin, Samuel G.; Morrow, Charles W.

2013-01-01T23:59:59.000Z

339

Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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.

Not Available

1992-10-01T23:59:59.000Z

340

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

SciTech Connect

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)

Hong, Dae-Seok; Kang, Il-Sik; Jang, Kyung-Duk; Jang, Won-Hyuk [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon (Korea, Republic of)] [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon (Korea, Republic of); Hoo, Wee-Teck [National Environment Agency, 40 Scotts Road 228231 (Singapore)] [National Environment Agency, 40 Scotts Road 228231 (Singapore)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Roles and effects of pyroprocessing for spent nuclear fuel management in South Korea  

Science Journals Connector (OSTI)

Abstract Republic of Korea (ROK) changed its spent nuclear fuel policy from the once-through usage and direct disposal to a total system approach that includes pyroprocessing, sodium-cooled fast reactors, and a two-tier geological repository to achieve a breakthrough for domestic deadlock situation and thus enable sustainable utilization of nuclear power, but caused disagreement in the bilateral negotiation with the United States (US) for the Nuclear Cooperation Agreement. Analysis has revealed that this shift is effective to make a breakthrough for domestic deadlock because it augments variety of technological options, with which more reversible decision-making process can be conducted to accommodate broad public needs. A trade-off has been explored first by deriving four engineering options from the ROK's system concept and then by comparing their performance from six viewpoints. The option including separation of high-heat emitting radionuclides by the electrolytic reduction process has been recommended. This option should be modified as exogenous and endogenous situations change in future. It is imperative for ROK to integrate a public-participatory decision-making process that works in concert with technology development. US can verify that ROK's motivation is not deviating from successful spent fuel management by checking if a transparent process with public participation is conducted.

Joonhong Ahn

2014-01-01T23:59:59.000Z

342

Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel  

SciTech Connect

The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses “throw-away” technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technology developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology.

Simpson, Michael F.; Benedict, Robert W.

2007-09-01T23:59:59.000Z

343

The Chemistry os Spent Nuclear Fuel From X-Ray Absorption Spectroscopy  

SciTech Connect

Present and future nuclear fuel cycles will require an understanding of the complex chemistry of trace fission products and transuranium actinides in spent nuclear fuel (SNF). Because of the unique analytical challenges presented by SNF to the materials scientist, many of its fundamental physical and chemical properties remain poorly understood, especially on the microscopic scale. Such an understanding of the chemical states of radionuclides in SNF would benefit development of technologies for fuel monitoring, fuel performance improvement and modeling, fuel reprocessing, and spent fuel storage and disposal. We have recently demonstrated the use of synchrotron x-ray absorption spectroscopy (XAS) to examine crystal chemical properties of actinides and fission products in extracted specimens of SNF. Information obtained includes oxidation state, chemical bond coordination, and quantitative elemental concentration and distribution. We have also used XAS in a scanning mode to obtain x-ray spectral micrographs with resolution approaching 1 micron. A brief overview of the technique will be presented, along with findings on uranium, plutonium, neptunium, technetium, and molybdenum in commercial PWR SNF specimens.

F.A. Fortner; A.J. Kropf; J.C. Cunnane

2006-09-21T23:59:59.000Z

344

Record of Decision for the Final EIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5091 5091 Friday May 17, 1996 Part IV Department of Energy Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel; Notice 25092 Federal Register / Vol. 61, No. 97 / Friday, May 17, 1996 / Notices DEPARTMENT OF ENERGY Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: DOE, in consultation with the Department of State, has decided to implement a new foreign research reactor spent fuel acceptance policy as specified in the Preferred Alternative contained in the Final Environmental Impact Statement on a Proposed

345

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

DOE Patents (OSTI)

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.

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

1987-11-05T23:59:59.000Z

346

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

DOE Patents (OSTI)

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.

Ackerman, John P. (Downers Grove, IL); Miller, William E. (Naperville, IL)

1989-01-01T23:59:59.000Z

347

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel  

SciTech Connect

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

1993-02-01T23:59:59.000Z

348

COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 2, User's manual  

SciTech Connect

COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code.

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

1986-11-01T23:59:59.000Z

349

Simulated Performance of the Integrated PNAR and SINRD Detector Designed for Spent Fuel Measurements at the Fugen Reactor in Japan  

SciTech Connect

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio was more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.

Lafleur, Adrienne M. [Los Alamos National Laboratory; Ulrich, Timothy J. II [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Japan Atomic Energy Agency; Bolind, Alan M. [Japan Atomic Energy Agency

2012-07-13T23:59:59.000Z

350

Measurement and analysis of gamma-rays emitted from spent nuclear fuel above 3 MeV  

SciTech Connect

The Next Generation Safeguard Initiative (NGSI) includes an effort to determine the mass content of fissile isotopes contained within spent fuel through the spectroscopy of high-energy delayed gamma rays. Studies being performed indicate the primary difficulty is the ability to detect the desired signal in the presence of the intense background associated with spent fuel fission products. An enabling technology for this application is high-resolution high-purity germanium (HPGe) detectors capable of operating efficiently in at extremely high count rates. This presentation will describe the prospects of high-rate germanium detectors and delayed-gamma techniques, primarily discussing the efforts to merge these into a unique and viable system for measuring spent fuel.

Rodriguez, Douglas C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Anderson, Elaina R. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); State Univ. of New York at Stony Brook, NY (United States); Anderson, Kevin K. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Campbell, Luke W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fast, James E. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Jarman, Kenneth D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Kulisek, Jonathan A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Orton, Christopher R. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Runkle, Robert C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Stave, Sean [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2013-08-28T23:59:59.000Z

351

COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method  

SciTech Connect

COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods.

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1986-11-01T23:59:59.000Z

352

CANDU core analysis with spent PWR fuel of fixed {sup 235}U and {sup 239}Pu content  

SciTech Connect

The direct use of spent pressurized water reactor (PWR) fuel in a CANDU (DUPIC) fuel cycle requires that the spent PWR fuel be refabricated as a DUPIC fuel to be used in a CANDU reactor that was originally designed for the natural uranium fuel. Because there is no separation of isotopes from the spent PWR fuel, the DUPIC fuel contains all the actinides and fission products as fuel constituents and, therefore, the isotopic composition changes depending on the initial and discharge conditions of PWR fuels. Such heterogeneity in fuel composition is not desirable, especially for a CANDU reactor that adopts an on-power refueling scheme, because the operational flexibility is reduced appreciably if the fuel bundles of different composition are loaded throughout the core.

Choi, Hangbok; Roh, Gyu H.; Park, J.W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-01T23:59:59.000Z

353

Characterization of the neutron source term and multiplicity of a spent fuel assembly in support of NSDA safeguards of spent nuclear fuel  

SciTech Connect

The gross neutron signal (GNS) is being considered as part of a fingerprinting or neutron balance approach to safeguards of spent nuclear fuel (SNF). Because the GNS is composed of many derivative components, understanding the time-dependent contribution of these derivative components is crucial to gauging the limitations of these approaches. The major components of the GNS are ({alpha}, n), spontaneous fission (SF), and multiplication neutrons. A methodology was developed to link MCNPX burnup output files to SOURCES4C input files for the purpose of automatically generating both the ({alpha}, n) and SF signals. Additional linking capabilities were developed to write MCNPX multiplication input files using the data obtained from the SOURCES4C output files. In this paper, the following are presented: (1) the relative contributions by source nuclide to the ({alpha}, n) signal as a function of initial enrichment/burnup/cooling time; (2) the relative contributions by source nuclide to the SF signal as a function of initial enrichment/burnup/cooling time; (3) the relative contributions by reaction type ({alpha},n vs. SF) to the GNS; and (4) the multiplication of the GNS as a function of initial enrichment/burnup/cooling time/counting environment. By developing these technologies to characterize the GNS, we can better evaluate the viability of the GNS fingerprint and neutron balance concepts for SNF.

Richard, Joshua G [Los Alamos National Laboratory; Fensin, Michael L [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Baciak, James [UNIV OF FL.

2010-01-01T23:59:59.000Z

354

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

SciTech Connect

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.

NONE

1995-07-01T23:59:59.000Z

355

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Transportation 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 Transportation of Spent Nuclear Fuel Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Task: Task: Task: Task: Identify Shortline Railroads Serving Nuclear Power Plants Identify Shortline Railroads Serving Nuclear Power Plants

356

Predictions of dry storage behavior of zircaloy clad spent fuel rods using deformation and fracture map analyses  

SciTech Connect

Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial dry storage temperatures of 420/sup 0/C for 1 year fuel cladding subjected to a constant stress of 70 MPa are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 850/sup 0/C stresses between 5 and 500 MPa. These maps are combined with a life fraction rule to predict the time to rupture of Zircaloy clad spent Light Water Reactor (LWR) fuel subjected to various storage conditions.

Tarn, J.C.L.; Madsen, N.H.; Chin, B.A.

1986-03-01T23:59:59.000Z

357

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank  

DOE Patents (OSTI)

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.

Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

358

Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario  

SciTech Connect

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.

Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

2007-01-01T23:59:59.000Z

359

Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel  

SciTech Connect

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility. Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack. (authors)

Masumi Wataru; Hisashi Kato; Satoshi Kudo; Naoko Oshima; Koji Wada [Central Research Institute of Electric Power Industry - CRIEPI (Japan); Hirofumi Narutaki [Electric Power Engineering Systems Co. Ltd. (Japan)

2006-07-01T23:59:59.000Z

360

Big Island Demonstration Project – Black Liquor  

Energy.gov (U.S. Department of Energy (DOE))

This fact sheet summarizes a U.S. Department of Energy Biomass Program research and development project.

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Black liquor gasification. Phase 2 final report  

SciTech Connect

The experimental work included 23 bench-scale tests in a 6-in.-diameter gasifier and two extended runs in a 33-in.-ID pilot-scale unit. The two pilot-scale runs included 26 test periods, each evaluated separately. The engineering analysis work consisted primarily of the correlation of test results and the development of a computer model describing the gasification process. 4 refs., 13 figs., 23 tabs.

Kohl, A.L.; Barclay, K.M.; Stewart, A.E.; Estes, G.R.

1984-11-28T23:59:59.000Z

362

DOE/EA-Ill7 ENVIRONMENTAL ASSESSMENT Management of Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Ill7 Ill7 ENVIRONMENTAL ASSESSMENT Management of Spent Nuclear Fuel on the Oak Ridge Reservation Oak Ridge, Tennessee February 1996 US. Department of Energy Oak Ridge Operations Oak Ridge, Tennessee DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. . DISCLAIMER i This report was prepared as an a m u n t 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 disclosed, or represents that its use would not infringe privately owned rights. Refer-

363

Brittle Failure Design Criteria for Ductile Cast Iron Spent-Fuel  

Office of Scientific and Technical Information (OSTI)

r r 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 disclosed, or represents that its use would not infringe privately owned rights. Refer- ence herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recom-

364

Alternative dispositioning methods for HEU spent nuclear fuel at the Savannah River Site  

SciTech Connect

The United States has a strong policy on prevention of the international spread of nuclear weapons. This policy was announced in Presidential Directive PDD-13 and summarized in a White House press release September 27, 1993. Two cornerstones of this policy are: seek to eliminate where possible the accumulation of stockpiles of highly- enriched uranium or plutonium; propose{hor_ellipsis}prohibiting the production of highly-enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside international safeguards. The Department of Energy is currently struggling to devise techniques that safely and efficiently dispose of spent nuclear fuel (SNF) while satisfying national non-proliferation policies. SRS plans and proposals for disposing of their SNF are safe and cost effective, and fully satisfy non-proliferation objectives.

Krupa, J.F.; McKibben, J.M.; Parks, P.B.; DuPont, M.E.

1995-11-01T23:59:59.000Z

365

Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage  

SciTech Connect

The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

Bailey, W.J.

1987-11-01T23:59:59.000Z

366

Solubility of Pu, Np, and U from Spent UO2?Fuel Under Inert/Reducing Conditions  

Science Journals Connector (OSTI)

The overall objective of this program is to improve the scientific understanding of processes that control the release of radioactive species especially actinides from spent fuel inside a disposal canister. The Swedish concept has focused on deep burial in the rock with an iron?lined Cu?canister. Corrosion of the canister iron insert will consume any residual oxygen and provide actively reducing conditions in any fluid phase. Therefore an investigation of the solubility of different radionuclides under actively reducing conditions) (Fe2+/H2) has been performed. The solubility of U Np and Pu is measured as a function of time for three different conditions: Ar atmosphere H2 atmosphere and H2 atmosphere with Fe(II) in solution.

Yngve Albinsson; Virginia Oversby; Arvid Ödegaard?Jensen; Lars Werme

2003-01-01T23:59:59.000Z

367

U.S. spent nuclear fuel management: Political, fiscal, and technical feasibility  

Science Journals Connector (OSTI)

Abstract Successful U.S. spent nuclear fuel management policy must satisfy political, fiscal, and technical constraints. Technical requirements have been thoroughly investigated in the United States and Nordic countries for volcanic tuff, salt, and granite. Fiscal planning requires an inflation-adjusted revenue stream and predictable real interest rate earnings on fund balances. A prompt solution satisfying political constraints requires compromise between the overlapping but distinct goals of seven different sets of interests at the federal level. Absent such compromise, there will be delay until sufficient support for one of three strategies evolves: (1) force the Yucca Mountain repository on Nevada, (2) open a centralized storage facility without coupling to repository licensing, or (3) follow a “consent-based” process for repository licensing. Formulations of each of these strategies to overcome impediments to their success are described.

Clifford Singer

2013-01-01T23:59:59.000Z

368

Calorimetric determination of the heat of combustion of spent Green River shale at 978 K  

SciTech Connect

A Calvet-type calorimeter was used to measure heats of combustion of spent Colorado oil shales. For Green River shale, the samples were members of a sink-float series spanning oil yields from 87 to 340 L . tonne/sup -1/. Shale samples (30-200 mg) are dropped into the calorimeter at high temperature, and a peak in the thermopile signal records the total enthalpy change of the sample between room temperature and the final temperature. Duplicate samples from the above sink-float series were first retorted at 773 K and then dropped separately into nitrogen and oxygen at 978 K. The resulting heats are subtracted to give the heat of combustion, and the results are compared to values from classical bomb calorimetry. The agreement shows that the heats of combustion of the organic component are well understood but that question remain on the reactions of the mineral components.

Mraw, S.C.; Keweshan, C.F.

1987-08-01T23:59:59.000Z

369

Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel  

SciTech Connect

This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

2004-04-01T23:59:59.000Z

370

Regulatory standards for permanent disposal of spent nuclear fuel and high-level radioactive waste.  

SciTech Connect

This paper provides a summary of observations drawn from twenty years of personal experience in working with regulatory criteria for the permanent disposal of radioactive waste for both the Waste Isolation Pilot Plant repository for transuranic defense waste and the proposed Yucca Mountain repository for spent nuclear fuel and high-level wastes. Rather than providing specific recommendations for regulatory criteria, my goal here is to provide a perspective on topics that are fundamental to how high-level radioactive waste disposal regulations have been implemented in the past. What are the main questions raised relevant to long-term disposal regulations? What has proven effective in the past? Where have regulatory requirements perhaps had unintended consequences? New regulations for radioactive waste disposal may prove necessary, but the drafting of these regulations may be premature until a broad range of policy issues are better addressed. In the interim, the perspective offered here may be helpful for framing policy discussions.

Swift, Peter N.

2010-08-01T23:59:59.000Z

371

Use of Integrated Decay Heat Limits to Facilitate Spent Nuclear Fuel Loading to Yucca Mountain  

SciTech Connect

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)

Li, Jun; Yim, Man-Sung; McNelis, David [Department of Nuclear Engineering, North Carolina State University (United States); Piet, Steven [Idaho National Laboratory (United States)

2007-07-01T23:59:59.000Z

372

Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements  

SciTech Connect

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of individual nuclides obtained from tandem EPRI-CELL/CINDER-2 calculations with those determined in documented benchmark inventory measurements of spent reactor fuel. The development of the popular /sup 148/Nd burnup measurement procedure is outlined, and areas of uncertainty in it and lack of clarity in its interpretation are indicated. Six inventory samples of varying quality and completeness are examined. The power histories used in the calculations have been listed for other users.

Wilson, W.B.; LaBauve, R.J.; England, T.R.

1982-01-01T23:59:59.000Z

373

Application of the RCP01 Code to Depletion of a PWR Spent Nuclear Fuel Sample  

SciTech Connect

An essential component of a proposed burnup credit methodology for commercial PWR spent nuclear fuel (SNF) is the validation of the tools used for isotopic and criticality calculations. A number of benchmark experiments have been analyzed to establish the validation of the tools and to determine biases and corrections. To benchmark the RCP01 Monte Carlo computer code, an isotopic validation study was conducted for one of the benchmark experiments, a SNF sample taken from the Calvert Cliffs PWR Unit-1 (CCPU1). Modeling considerations and nuclear data associated with the RCP01 transport/depletion calculations are discussed. The accuracy of RCP01 calculations is demonstrated to be very good when RCP01 results are compared to destructive chemical assay data for major actinides and important fission products in the SNF sample.

Joo, Hansem

2002-01-01T23:59:59.000Z

374

Creep strains predicted from constitutive equations for Zircaloy-clad spent fuel rods  

SciTech Connect

Integrity of a high percentage of the Zircaloy (zirconium-alloy) clad spent fuel rods during extended dry storage is a design consideration of the storage system. Maintaining cladding integrity after permanent disposal placement may also be desired, although other barriers will be engineered to contain the radioactive waste. The limits to the temperatures the Zircaloy tubes can sustain over extended times are principally determined by creep at stresses (from internal gas pressure) and temperatures relevant to dry storage conditions. Excessive creep may lead to an eventual ductile fracture crack. The thermal design criteria assure compatibility of the heat load, the heat transfer properties, and the materials response. Therefore, the ability to predict the long-term integrity of Zircaloy is important to the design requirements. To do this reliably requires both a theoretical and empirical description of creep deformation and fracture.

Rosen, R.S.; O`Connell, W.J.

1994-11-01T23:59:59.000Z

375

Application of zircaloy deformation and fracture maps to predicting dry spent fuel storage conditions  

SciTech Connect

The research reported details the development of theoretical deformation and fracture maps for Zircaloy at temperatures between 50 and 850/sup 0/C and for stresses between 5 and 500 MPa. The maps predict behavior which is in agreement with a large number of experimental observations reported in the literature. Transformations in deformation and fracture mechanisms, creep rate magnitudes, and times to fracture are predicted from the maps. Results are reported for grain sizes of 5 ..mu..m. The deformation and fracture maps are altered for irradiation effects based upon high stress-high temperature, short time, postirradiation data. Maximum allowable temperature limits to achieve 40 years dry storage are calculated as a function of time spent in wet pool storage using a decaying cask temperature function and a life fraction rule. 122 references, 15 figures, 4 tables.

Chin, G.A.; Madsen, N.H.; Khan, M.A.

1983-08-01T23:59:59.000Z

376

Potential application of Zircaloy chemical embrittlement to volume reduction of spent-fuel cladding  

SciTech Connect

Embrittlement of Zircaloy fuel cladding tubes by corrosion media was studied from the viewpoint of its applicability to spent-fuel reprocessing. The results from irradiated as well as unirradiated tubes are summarized as follows: (1) when iodine was employed as the solute, the use of methanol as the solvent caused significant embrittlement of the Zircaloy; (2) for the iodine-methanol solution, the embrittlement increased with the iodine content but saturated at 1 wt%; (3) a water content of up to 10 vol% in the iodine-methanol solution did not decrease the extent of embrittlement; (4) fracture was of the grain-boundary type, and a fuel cladding tube irradiated to [approximately] 35GWd/t showed the same embrittlement behavior as an unirradiated one.

Nakatsuka, Masafumi (Nippon Nuclear Fuel Development Co., Ltd., Ibaraki (Japan))

1993-09-01T23:59:59.000Z

377

Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns  

SciTech Connect

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 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.

Marsha Keister; Kathryn McBride

2006-08-01T23:59:59.000Z

378

Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor  

SciTech Connect

A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

Steven D. Herrmann; Michael F. Simpson

2006-08-01T23:59:59.000Z

379

Container for reprocessing and permanent storage of spent nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Forsberg, C.W.

1992-03-24T23:59:59.000Z

380

United States Program on Spent Nuclear Fuel and High-Level Radioactive Waste Management  

SciTech Connect

The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas.

Stewart, L.

2004-10-03T23:59:59.000Z

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

DOCUMENTATION OF NATIONAL WEATHER CONDITIONS AFFECTING LONG-TERM DEGRADATION OF COMMERCIAL SPENT NUCLEAR FUEL AND DOE SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE  

SciTech Connect

The U.S. Department of Energy (DOE) is preparing a proposal to construct, operate 2nd monitor, and eventually close a repository at Yucca Mountain in Nye County, Nevada, for the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). As part of this effort, DOE has prepared a viability assessment and an assessment of potential consequences that may exist if the repository is not constructed. The assessment of potential consequences if the repository is not constructed assumes that all SNF and HLW would be left at the generator sites. These include 72 commercial generator sites (three commercial facility pairs--Salem and Hope Creek, Fitzpatrick and Nine Mile Point, and Dresden and Morris--would share common storage due to their close proximity to each other) and five DOE sites across the country. DOE analyzed the environmental consequences of the effects of the continued storage of these materials at these sites in a report titled Continued Storage Analysis Report (CSAR; Reference 1 ) . The CSAR analysis includes a discussion of the degradation of these materials when exposed to the environment. This document describes the environmental parameters that influence the degradation analyzed in the CSAR. These include temperature, relative humidity, precipitation chemistry (pH and chemical composition), annual precipitation rates, annual number of rain-days, and annual freeze/thaw cycles. The document also tabulates weather conditions for each storage site, evaluates the degradation of concrete storage modules and vaults in different regions of the country, and provides a thermal analysis of commercial SNF in storage.

W. L. Poe, Jr.; P.F. Wise

1998-11-01T23:59:59.000Z

382

DOE Spent Nuclear Fuel Information In Support of TSPA-VA  

SciTech Connect

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.

A. Brewer; D. Cresap; D. Fillmore; H. Loo; M. Ebner; R. McCormack

1998-09-01T23:59:59.000Z

383

Analysis of spent ion exchange media: Superlig 639 and Superlig 644  

SciTech Connect

The current BNFL Inc. flowsheet for the pretreatment of the Hanford High-Level tank wastes includes the use of Superlig{reg_sign} materials for the removal of {sup 137}Cs and {sup 99}Tc from the aqueous fraction of the waste. The cesium-selective Superlig{reg_sign} 644 (SL-644) and the technetium-selective Superlig{reg_sign} 639 (SL-639) have been evaluated in tests with actual waste samples. These materials have a finite processing lifetime in the plant and will need to be disposed of. The composition and level of residual radionuclide contamination is important for assessing various disposal pathways for the Superlig{reg_sign} materials. This report contains the results of analyses of subsamples of the SL-639 and SL 644 materials that have been used in small column testing of actual waste samples at the Radiochemical Processing Laboratory. The wastes that have been tested include samples from Tanks 241-AW-101 and 241-AN-107. The analyses of the spent resins include inductively coupled plasma/atomic emission spectrometry (ICP-AES) for metals, cold vapor atomic absorption (CVAA) spectroscopy for mercury, gamma energy analysis (GEA) for radionuclides and inductively coupled plasma/mass spectrometry (ICP-MS) for selected metals and radionuclides. While these results provide an indication of the analyte concentrations that may be left on the spent resin, they do not fully represent the concentrations that may be found after extended plant processing with additional load/elute cycles and different waste compositions. BNFL estimates that the SL-644 may last for 100 load/elute cycles with Envelope A and C wastes and 20 cycles with Envelope B wastes. The number of useable load/elute cycles for the SL-639 is not well defined, but is likely on the order of hundreds.

DE Kurath; JJ Wagner

2000-05-31T23:59:59.000Z

384

Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel  

SciTech Connect

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.

Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Goluoglu, Sedat [ORNL; Hollenbach, Daniel F [ORNL; Fox, Patricia B [ORNL

2007-10-01T23:59:59.000Z

385

Evaluation of Nondestructive Assay/Nondestructive Examination Capabilities for Department of Energy Spent Nuclear Fuel  

SciTech Connect

This report summarizes an evaluation of the potential use of nondestructive assay (NDA) and nondestructive examination (NDE) technologies on DOE spent nuclear fuel (SNF). It presents the NDA/NDE information necessary for the National Spent Nuclear Fuel Program (NSNFP) and the SNF storage sites to use when defining that role, if any, of NDA/NDE in characterization and certification processes. Note that the potential role for NDA/NDE includes confirmatory testing on a sampling basis and is not restricted to use as a primary, item-specific, data collection method. The evaluation does not attempt to serve as a basis for selecting systems for development or deployment. Information was collected on 27 systems being developed at eight DOE locations. The systems considered are developed to some degree, but are not ready for deployment on the full range of DOE SNF and still require additional development. The system development may only involve demonstrating performance on additional SNF, packaging the system for deployment, and developing calibration standards, or it may be as extensive as performing additional basic research. Development time is considered to range from one to four years. We conclude that NDA/NDE systems are capable of playing a key role in the characterization and certification of DOE SNF, either as the primary data source or as a confirmatory test. NDA/NDE systems will be able to measure seven of the nine key SNF properties and to derive data for the two key properties not measured directly. The anticipated performance goals of these key properties are considered achievable except for enrichment measurements on fuels near 20% enrichment. NDA/NDE systems can likely be developed to measure the standard canisters now being considered for co-disposal of DOE SNF. This ability would allow the preparation of DOE SNF for storage now and the characterization and certification to be finalize later.

Luptak, A.J.; Bulmahn, K.D.

1998-09-01T23:59:59.000Z

386

Overview of Requirements for Using Overweight Vehicles to Ship Spent Nuclear Fuel  

SciTech Connect

The U.S. Department of Energy's (DOE'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, considered a range of options for transportation. In evaluating the impacts of the mostly-legal weight truck scenario, DOE assumed that some shipments would use overweight trucks. The use of overweight trucks is also considered in the Draft Supplemental 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, issued for public comment in Fall 2007. With the exception of permit requirements and operating restrictions, the vehicles for overweight shipments would be similar to legal-weight truck shipments but might weigh as much as 52,200 kilograms (115,000 pounds). The use of overweight trucks was determined to be acceptable for the Office of Civilian Radioactive Waste Management (OCRWM) Program because the payload is not divisible and the packaging alone may make shipments overweight. Overweight truck shipments are common, and states routinely issue overweight permits, some for vehicles with a gross vehicle weight up to 58,500 kilograms (129,000 pounds). This paper will present an overview of state overweight truck permitting policies and national and regional approaches to promote safety and uniformity. In conclusion: Overweight truck shipments are made routinely by carriers throughout the country. State permits are obtained by the carriers or by companies that provide permitting services to the carriers. While varying state permit restrictions may add complexity to OCRWM's planning activities, the well-established experience of commercial carriers and efforts to bring uniformity to the permitting process should allow the overweight shipment of SNF to be a viable option. (authors)

Thrower, A.W. [U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, DC (United States); Offner, J. [Booz Allen Hamilton, Washington, DC (United States); Bolton, P. [Booz Allen Hamilton, Santa Fe, NM (United States)

2008-07-01T23:59:59.000Z

387

However, the years that I spent in Egypt as a child and young adult blessed me with a solid foundation  

E-Print Network (OSTI)

#12;However, the years that I spent in Egypt as a child and young adult blessed me with a solid of the world that often left me distressed. Many thought 1rode camels all day long in Egypt. The truth is that 1had never ridden a camel in Egypt. Many also imagined that most of my fellow Egyptians drilled

Zewail, Ahmed

388

Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes  

SciTech Connect

The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage.

Harmon, K.M.; Johnson, A.B. Jr.

1984-04-01T23:59:59.000Z

389

COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel  

SciTech Connect

The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37/sup 0/C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8/sup 0/C (100/sup 0/F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance.

Rector, D.R.; Wheeler, C.L.

1986-01-01T23:59:59.000Z

390

Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste  

Energy.gov (U.S. Department of Energy (DOE))

[In Support of a Comprehensive National Nuclear Fuel Cycle Strategy, Volumes I and II (Appendices)] 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 isolation.

391

Development of U-Frame Bending System for Studying the Vibration Integrity of Spent Nuclear Fuel  

SciTech Connect

A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. Design and analysis, fabrication, modification, calibration, and instrumentation are described. The system is composed of a U-frame testing setup for imposing bending loads on the spent fuel rod test specimen and a method for measuring the curvature of the rod during bending. The U-frame setup consists of two rigid arms, linking members, and linkages to a universal testing machine. The test specimen s curvature of bending is obtained through a three-point deflection measurement method consisting of three LVDTs mounted to the side connecting plates of the U-frame to capture the deformation of the test specimen. The system has some unique features: 1) The test specimen is installed by simple insertion using linear bearings incorporated with rigid sleeves. 2) Reverse cyclic bending tests can be carried out effectively and efficiently by push and pull at the loading point of the setup. Any test machine with a linear motion function can be used to drive the setup. 3) The embedded and preloaded linear roller bearings eliminate the backlash that exists in the conventional reverse bend tests. 4) The number of linkages between the U-frame and the universal machine is minimized. Namely, there are only two linkages needed at the two loading points of a U-frame setup, whereas a conventional four/three-point bend test frame requires four linkages. 5) The curvature measurement is immune to the effects arising from compliant layers and the rigid body motion of the machine. The compliant layers are used at the holding areas of the specimen to prevent contact damage. The tests using surrogate specimens composed of SS cladding/tube revealed several important phenomena that may cast light on the expected response of a spent fuel rod: 1) Cyclic quasi-static load (10 N/s under force control) in compressive mode (with respect to that at the loading point of the U-frame) produced increased irreversible or plastic curvature and also increased flexural rigidity of the surrogate rod. 2) Dynamic cyclic load (at least 1 Hz) in compressive mode resulted in increased flexural rigidity of the surrogate rod prior to SS cladding fracture. 3) Pellets and epoxy bonding exhibited various effects on the response of surrogate rods during the loading process as validated from static tests. 4) Dynamic cyclic load (2 Hz) in reverse mode demonstrated a substantial cyclic softening before the fracture of the surrogate rod. The degree of decrease in flexural rigidity was consistent in both measurement and on-line monitoring. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

Wang, Hong [ORNL; Wang, Jy-An John [ORNL; Tan, Ting [ORNL; Jiang, Hao [ORNL; Cox, Thomas S [ORNL; Howard, Rob L [ORNL; Bevard, Bruce Balkcom [ORNL; Flanagan, Michelle E [ORNL

2013-01-01T23:59:59.000Z

392

Chemical speciation of neptunium in spent fuel. Annual report for period 15 August 1999 to 15 August 2000  

SciTech Connect

(B204) This project will examine the chemical speciation of neptunium in spent nuclear fuel. The R&D fields covered by the project include waste host materials and actinide chemistry. Examination of neptunium is chosen since it was identified as a radionuclide of concern by the NERI workshop. Additionally, information on the chemical form of neptunium in spent fuel is lacking. The identification of the neptunium species in spent fuel would allow a greater scientific based understanding of its long-term fate and behavior in waste forms. Research to establish the application and development of X-ray synchrotrons radiation (XSR) techniques to determine the structure of aqueous, adsorbed, and solid actinide species of importance to nuclear considerations is being conducted at Argonne. These studies extend current efforts within the Chemical Technology Division at Argonne National Laboratory to investigate actinide speciation with more conventional spectroscopic and solids characterization (e.g. SEM, TEM, and XRD) methods. Our project will utilize all these techniques for determining neptunium speciation in spent fuel. We intend to determine the chemical species and oxidation state of neptunium in spent fuel and alteration phases. Different types of spent fuel will be examined. Once characterized, the chemical behavior of the identified neptunium species will be evaluated if it is not present in the literature. Special attention will be given to the behavior of the neptunium species under typical repository near-field conditions (elevated temperature, high pH, varying Eh). This will permit a timely inclusion of project results into near-field geochemical models. Additionally, project results and methodologies have applications to neptunium in the environment, or treatment of neptunium containing waste. Another important aspect of this project is the close cooperation between a university and a national laboratory. The PI has a transuranic laboratory at MIT where students can perform spectroscopic and radiochemical experiments. Through the ANL partner, students can have additional experience performing research in a DOE setting. This will provide a unique and constructive opportunity for developing quality graduate students with experience and expertise in handling actinides. Our ability to produce experienced actinide scientists is currently restricted by the dearth of radiochemistry and nuclear research at universities. Regardless of all else, future researchers must be trained and educated if the United States is to maintain a leadership role in nuclear technology. This project provides such an opportunity.

Ken Czerwinski; Don Reed

2000-09-01T23:59:59.000Z

393

REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES  

SciTech Connect

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.

W. Lee Poe, Jr

1998-10-01T23:59:59.000Z

394

Laboratory study on the behaviour of spent AA household alkaline batteries in incineration  

SciTech Connect

The quantitative evaluation of emissions from incineration is essential when Life Cycle Assessment (LCA) studies consider this process as an end-of-life solution for some wastes. Thus, the objective of this work is to quantify the main gaseous emissions produced when spent AA alkaline batteries are incinerated. With this aim, batteries were kept for 1 h at 1273 K in a refractory steel tube hold in a horizontal electric furnace with temperature control. At one end of the refractory steel tube, a constant air flow input assures the presence of oxygen in the atmosphere and guides the gaseous emissions to a filter system followed by a set of two bubbler flasks having an aqueous solution of 10% (v/v) nitric acid. After each set of experiments, sulphur, chlorides and metals (As, Cd, Co, Cr, Cu, Fe, Hg, Mn, Ni, Pb, Sb, Tl and Zn) were analyzed in both the solutions obtained from the steel tube washing and from the bubblers. Sulphur, chlorides and metals were quantified, respectively, using barium sulfate gravimetry, the Volhard method and atomic absorption spectrometry (AAS). The emissions of zinc, the most emitted metal, represent about 6.5% of the zinc content in the batteries. Emissions of manganese (whose oxide is the main component of the cathode) and iron (from the cathode collector) are negligible when compared with their amount in AA alkaline batteries. Mercury is the metal with higher volatility in the composition of the batteries and was collected even in the second bubbler flask. The amount of chlorides collected corresponds to about 36% of the chlorine in the battery sleeve that is made from PVC. A considerable part of the HCl formed in PVC plastic sleeve incineration is neutralized with KOH, zinc and manganese oxides and, thus, it is not totally released in the gas. Some of the emissions are predictable through a thermodynamic data analysis at temperatures in the range of 1200-1300 K taking into account the composition of the batteries. This analysis was done for most of potential reactions between components in the batteries as well as between them and the surrounding atmosphere and it reasonably agrees the experimental results. The results obtained show the role of alkaline batteries at the acid gases cleaning process, through the neutralization reactions of some of their components. Therefore, LCA of spent AA alkaline batteries at the municipal solid waste (MSW) incineration process must consider this contribution.

Almeida, Manuel F. [LEPAE, Faculty of Engineering, University of Porto, Rua Dr. Roberto Frias, 4200-465 Porto (Portugal)], E-mail: mfa@fe.up.pt; Xara, Susana M.; Delgado, Julanda; Costa, Carlos A. [LEPAE, Faculty of Engineering, University of Porto, Rua Dr. Roberto Frias, 4200-465 Porto (Portugal)

2009-01-15T23:59:59.000Z

395

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES  

SciTech Connect

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

DOE

1997-04-01T23:59:59.000Z

396

TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY  

SciTech Connect

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study is to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance. Technical challenges include pure bending implementation, remote installation and detachment of the SNF test specimen, test specimen deformation measurement, and identification of a driving system suitable for use in a hot cell. Surrogate test specimens have been used to calibrate the test setup and conduct systematic cyclic tests. The calibration and systematic cyclic tests have been used to identify test protocol issues prior to implementation in the hot cell. In addition, cyclic hardening in unidirectional bending and softening in reverse bending were observed in the surrogate test specimens. The interface bonding between the surrogate clad and pellets was found to impact the bending response of the surrogate rods; confirming this behavior in the actual spent fuel segments will be an important aspect of the hot cell test implementation,

Wang, Jy-An John [ORNL] [ORNL; Wang, Hong [ORNL] [ORNL; Bevard, Bruce Balkcom [ORNL] [ORNL; Howard, Rob L [ORNL] [ORNL; Flanagan, Michelle [U.S. Nuclear Regulatory Commission] [U.S. Nuclear Regulatory Commission

2013-01-01T23:59:59.000Z

397

Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation  

SciTech Connect

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.

Wang, Jy-An John [ORNL] [ORNL; Wang, Hong [ORNL] [ORNL; Bevard, Bruce Balkcom [ORNL] [ORNL; Howard, Rob L [ORNL] [ORNL; Flanagan, Michelle [U.S. Nuclear Regulatory Commission] [U.S. Nuclear Regulatory Commission

2013-01-01T23:59:59.000Z

398

Recovery of minor actinides from spent fuel using TPEN-immobilized gels  

SciTech Connect

A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (so called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by changing the pH of the eluent to 2.0. These results These results prove that the proposed recovery process for MAs is a potential candidate for future reprocessing methods based on the extraction chromatographic technique. (authors)

Koyama, S.; Suto, M.; Ohbayashi, H. [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai (Japan); Oaki, H. [Solutions Research Organization, Tokyo Institute of Technology, Tokyo (Japan); Takeshita, K. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

2013-07-01T23:59:59.000Z

399

Simplified Reference Electrode for Electrorefining of Spent Nuclear Fuel in High Temperature Molten Salt  

SciTech Connect

Pyrochemical processing plays an important role in development of proliferation- resistant nuclear fuel cycles. At the Idaho National Laboratory (INL), a pyrochemical process has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor II (EBR-II) in the last decade. Electrorefining in a high temperature molten salt is considered a signature or central technology in pyroprocessing fuel cycles. Separation of actinides from fission products is being demonstrated by electrorefining the spent fuel in a molten UCl3-LiCl-KCl electrolyte in two engineering scale electrorefiners (ERs). The electrorefining process is current controlled. The reference electrode provides process information through monitoring of the voltage difference between the reference and the anode and cathode electrodes. This information is essential for monitoring the reactions occurring at the electrodes, investigating separation efficiency, controlling the process rate, and determining the process end-point. The original reference electrode has provided good life expectancy and signal stability, but is not easily replaceable. The reference electrode used a vycor-glass ion-permeable membrane containing a high purity silver wire with one end positioned in ~2 grams of LiCl/KCl salt electrolyte with a low concentration (~1%) AgCl. It was, however, a complex assembly requiring specialized skill and talent to fabricate. The construction involved multiple small pieces, glass joints, ceramic to glass joints, and ceramic to metal joints all assembled in a high purity inert gas environment. As original electrodes reached end-of-life it was uncertain if the skills and knowledge were readily available to successfully fabricate replacements. Experimental work has been conducted to identify a simpler electrode design while retaining the needed long life and signal stability. This improved design, based on an ion-permeable membrane of mullite has been completed. Use of the silver wire and electrolyte composition remains the same. The resulting design maximizes the use of commercial components and can be fabricated with commonly available skills. This has resulted in a significant reduction of effort and cost to fabricate replacements. The piece count requiring assembly in a high purity glove box atmosphere has been reduced by over half and all specialized joints have been eliminated. The new design has been tested in a lab scale electrorefiner and has also been successfully scaled up and installed in the engineering scale electrorefiners.

Kim Davies; Shelly X Li

2007-09-01T23:59:59.000Z

400

DOE/EIS-0279; Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement (March 2000)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

March 2000 Summary S-iii COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Savannah River Site, Spent Nuclear Fuel Management Final Environmental Impact Statement (DOE/EIS-0279) CONTACT: For additional information on this environmental impact statement, write or call: Andrew R. Grainger, NEPA Compliance Officer U.S. Department of Energy, Savannah River Operations Office, Building 742A, Room 183 Aiken, South Carolina 29802 Attention: Spent Nuclear Fuel Management EIS Local and Nationwide Telephone: (800) 881-7292 Email: nepa@SRS.gov The EIS is also available on the internet at: http://tis.eh.doe.gov/nepa/docs/docs.htm. For general information on the process that DOE follows in complying with the National Environmental Policy Act, write or call:

Note: This page contains sample records for the topic "liquor spent sulfite" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

DOE/EIS-0279; Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement (March 2000)  

NLE Websites -- All DOE Office Websites (Extended Search)

Summary Summary S-iii COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Savannah River Site, Spent Nuclear Fuel Management Final Environmental Impact Statement (DOE/EIS-0279) CONTACT: For additional information on this environmental impact statement, write or call: Andrew R. Grainger, NEPA Compliance Officer U.S. Department of Energy, Savannah River Operations Office, Building 742A, Room 183 Aiken, South Carolina 29802 Attention: Spent Nuclear Fuel Management EIS Local and Nationwide Telephone: (800) 881-7292 Email: nepa@SRS.gov The EIS is also available on the internet at: http://tis.eh.doe.gov/nepa/docs/docs.htm. For general information on the process that DOE follows in complying with the National Environmental Policy Act, write or call:

402

DOE/EIS-0279; Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement (March 2000)  

NLE Websites -- All DOE Office Websites (Extended Search)

Cover Sheet Cover Sheet iii COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Savannah River Site, Spent Nuclear Fuel Management Final Environmental Impact Statement (DOE/EIS-0279) CONTACT: For additional information on this environmental impact statement, write or call: Andrew R. Grainger, NEPA Compliance Officer U.S. Department of Energy, Savannah River Operations Office, Building 742A, Room 183 Aiken, South Carolina 29802 Attention: Spent Nuclear Fuel Management EIS Local and Nationwide Telephone: (800) 881-7292 Email: nepa@SRS.gov The EIS is also available on the internet at: http://tis.eh.doe.gov/nepa/docs/docs.htm. For general information on the process that DOE follows in complying with the National Environmental Policy Act, write or call:

403

Record of Decision for a Dry Storage Container System for the Management of Navel Spent Nuclear Fuel, January 7, 1997  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5 5 Federal Register / Vol. 62, No. 5 / Wednesday, January 8, 1997 / Notices Department of the Navy Record of Decision for a Dry Storage Container System for the Management of Naval Spent Nuclear Fuel SUMMARY: Pursuant to section 102(2) of the National Environmental Policy Act (NEPA) of 1969; the Council on Environmental Quality regulations implementing NEPA procedures, 40 CFR Parts 1500-1508; and Chief of Naval Operations Environmental and Natural Resources Program Manual, OPNAV Instruction 5090.1B; the Department of the Navy announces its decision to implement the preferred alternative (dual-purpose canisters) identified in the final Environmental Impact Statement for a Container System for the Management of Naval Spent Nuclear Fuel (EIS) dated November 1996. The Department of

404

Draft Environmental Impact Statement for a Container System for the Management of Naval Spent Nuclear Fuel EIS-0251  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Document ID 51 Document ID 51 Commenter: Daniel Nix - Western Interstate Energy Board, Colorado Response to Comment: A. The Navy extended the comment period from 45 to 60 days (ending July 18, 1996) in response to requests from the state of Nevada. A further extension could not be provided because of the need to complete the EIS to support actions required under a court agreement among the Department of Energy, Navy, and State of Idaho covering spent fuel management at the Idaho National Engineering Laboratory. B.&D. The Board's comment is correct that the EIS is limited to naval spent nuclear fuel and Navy- generated special case waste. The Board's comment is incorrect in the implication that transportation to Yucca Mountain is supported by the EIS. The proposed action of this EIS

405

Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies  

SciTech Connect

A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application.

Ham, Y S; Maldonado, G I; Burdo, J; He, T

2006-10-10T23:59:59.000Z

406

Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE))

The Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel report assesses the technical options for the safe and permanent disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. Specifically, it considers whether DOE-managed HLW and SNF should be disposed of with commercial SNF and HLW in one geologic repository or whether there are advantages to developing separate geologic disposal pathways for some DOE-managed HLW and SNF. The report recommends that the Department begin implementation of a phased, adaptive, and consent-based strategy with development of a separate mined repository for some DOE-managed HLW and cooler DOE-managed SNF.

407

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF  

SciTech Connect

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.

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

2014-03-31T23:59:59.000Z

408

Spent fuel test-climax: a test of geologic storage of high-level waste in granite  

SciTech Connect

A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented.

Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

1981-01-01T23:59:59.000Z

409

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost  

SciTech Connect

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC ({approx}49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

410

U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3  

Gasoline and Diesel Fuel Update (EIA)

7 7 Table 1 | Table 2 Table 3. Annual Spent Fuel Discharges and Burnup, 1968 - 2002 Year Number of Assemblies a Initial Uranium Content (Metric Tonnes of Uranium) Average Burnup (GWDt/ MTU) All Discharged Assemblies BWR PWR HTGR Total BWR PWR HTGR Total BWR PWR HTGR 1968 5 0 0 5 0.6 0.6 1.7 1969 97 0 0 97 9.9 9.9 16.6 1970 29 99 0 128 5.6 39.0 44.6 0.3 18.4 1971 413 113 0 526 64.7 44.5 109.2 8.3 23.8 1972 801 282 0 1,083 145.8 99.9 245.7 7.1 22.1 1973 564 165 0 729 93.5 67.1 160.6 13.2 24.2 1974 1,290 575 0 1,865 241.6 207.7 449.3 13.1 18.4 1975 1,223 797 0 2,020 225.9 321.7 547.6 17.1 18.2 1976 1,666 931 0 2,597 298.1 401.0 699.1 13.6 22.4

411

Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect

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.

NONE

1997-12-01T23:59:59.000Z

412

MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4  

SciTech Connect

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 le

Carbajo, Juan J [ORNL] [ORNL

2012-01-01T23:59:59.000Z

413

Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses  

SciTech Connect

This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions.

Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

1986-12-01T23:59:59.000Z

414

Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation  

SciTech Connect

The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF for transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)

Svitak, F.; Broz, V.; Hrehor, M.; Marek, M.; Novosad, P.; Podlaha, J.; Rychecky, J. [Nuclear Research Institute Rez plc, Husinec 130, CZ-25068 (Czech Republic)

2008-07-15T23:59:59.000Z

415

Technical Cross-Cutting Issues for the Next Generation Safeguards Initiative's Spent Fuel Nondestructive Assay Project  

SciTech Connect

Ever since there has been spent fuel (SF), researchers have made nondestructive assay (NDA) measurements of that fuel to learn about its content. In general these measurements have focused on the simplest signatures (passive photon and total neutron emission) and the analysis has often focused on diversion detection and on determining properties such as burnup (BU) and cooling time (CT). Because of shortcomings in current analysis methods, inspectorates and policy makers are interested in improving the state-of-the-art in SF NDA. For this reason the U.S. Department of Energy, through the Next Generation Safeguards Initiative (NGSI), targeted the determination of elemental Pu mass in SF as a technical goal. As part of this research effort, 14 nondestructive assay techniques were studied . This wide range of techniques was selected to allow flexibility for the various needs of the safeguards inspectorates and to prepare for the likely integration of one or more techniques having complementary features. In the course of researching this broad range of NDA techniques, several cross-cutting issues were. This paper will describe some common issues and insights. In particular we will describe the following: (1) the role of neutron absorbers with emphasis on how these absorbers vary in SF as a function of initial enrichment, BU and CT; (2) the need to partition the measured signal among different isotopic sources; and (3) the importance of the “first generation” concept which indicates the spatial location from which the signal originates as well as the isotopic origins.

Tobin, S. J.; Menlove, H. O.; Swinhoe, Martyn T.; Blanc, P.; Burr, T.; Evans, L. G.; Favalli, A.; Fensin, M. L.; Freeman, C. R.; Galloway, J.; Gerhart, J.; Rajasingam, A.; Rauch, E.; Sandoval, N. P.; Trellue, H.; Ulrich, T. J.; Conlin, J. L.; Croft, S.; Hendricks, John; Henzl, V.; Henzlova, D.; Eigenbrodt, J. M.; Koehler, W. E.; Lee, D. W.; Lee, T. H.; Lafleur, A. M.; Schear, M. A.; Humphrey, M. A.; Smith, Leon E.; Anderson, Kevin K.; Campbell, Luke W.; Casella, Andrew M.; Gesh, Christopher J.; Shaver, Mark W.; Misner, Alex C.; Amber, S. D.; Ludewigt, Bernhard A.; Quiter, B.; Solodov, Alexander; Charlton, W.; Stafford, A.; Romano, C.; Cheatham, J.; Ehinger, Michael; Thompson, S. J.; Chichester, David; Sterbentz, James; Hu, Jianwei; Hunt, A.; Mozin, Vladimir V.; Richard, J. G.

2012-03-01T23:59:59.000Z

416

Characterization of LWR spent fuel MCC-approved testing material-ATM-101  

SciTech Connect

The characterization data, obtained to date, for Materials Characterization Center (MCC) Approved Testing Materials (ATM)-101, spent fuel from H.B. Robinson, Unit 2, Assembly BO-5, are described. ATM-101 consists of 27 equal-length segments from nine fuel rods. Characterizations provided for ATM-101 include, (1) reactor, assembly, and fuel rod descriptions, (2) Assembly BO-5 irradiation history, (3) a description of unusual incidents that occurred to the rods, (4) fission gas release measurements, (5) results of ceramography/metallography examinations, (6) fuel burnup measurement results and correlations, (7) results of gamma scanning, (8) calculated values of the radionuclide inventory, and (9) results of a radionuclide chemical overcheck. Calculations for and measurement of radial distributions of selected radionuclides are planned. A description of pertinent results from other studies on sibling rods from Assembly BO-5 is also included. The distribution of ATM-101 to date is described along with characterization results on specially processed material. It is intended that this report be revised and updated as additional characterization data become available. 6 references, 23 figures, 19 tables.

Barner, J.O.

1984-06-01T23:59:59.000Z

417

US Department of Energy Storage of Spent Fuel and High Level Waste  

SciTech Connect

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.

Sandra M Birk

2010-10-01T23:59:59.000Z

418

Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX  

SciTech Connect

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)

Van den Akker, B.P.; Ahn, J. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720 (United States)

2013-07-01T23:59:59.000Z

419

DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel  

SciTech Connect

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 storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all 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 (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) 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. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. 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.

Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1995-11-30T23:59:59.000Z