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Note: This page contains sample records for the topic "high activity waste" from the National Library of EnergyBeta (NLEBeta).
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1

WASTE SOLIDIFICATION BUILDING BENCH SCALE HIGH ACTIVITY WASTE SIMULANT VARIABILITY STUDY FY2008  

SciTech Connect (OSTI)

The primary objective of this task was to perform a variability study of the high activity waste (HAW) acidic feed to determine the impact of feed variability on the quality of the final grout and on the mixability of the salt solution into the dry powders. The HAW acidic feeds were processed through the neutralization/pH process, targeting a final pH of 12. These fluids were then blended with the dry materials to make the final waste forms. A secondary objective was to determine if elemental substitution for cost prohibitive or toxic elements in the simulant affects the mixing response, thus providing a more economical simulant for use in full scale tests. Though not an objective, the HAW simulant used in the full scale tests was also tested and compared to the results from this task. A statistically designed test matrix was developed based on the maximum molarity inputs used to make the acidic solutions. The maximum molarity inputs were: 7.39 HNO{sub 3}, 0.11618 gallium, 0.5423 silver, and 1.1032 'other' metals based on their NO{sub 3}{sup -} contribution. Substitution of the elements aluminum for gallium and copper for silver was also considered in this test matrix, resulting in a total of 40 tests. During the NaOH addition, the neutralization/pH adjustment process was controlled to a maximum temperature of 60 C. The neutralized/pH adjusted simulants were blended with Portland cement and zircon flour at a water to cement mass ratio of 0.30. The mass ratio of zircon flour to Portland cement was 1/12. The grout was made using a Hobart N-50 mixer running at low speed for two minutes to incorporate and properly wet the dry solids with liquid and at medium speed for five minutes for mixing. The resulting fresh grout was measured for three consecutive yield stress measurements. The cured grout was measured for set, bleed, and density. Given the conditions of preparing the grout in this task, all of the grouts were visually well mixed prior to preparing the grouts for measurements. All of the cured grouts were measured for bleed and set. All of the cured grouts satisfied the bleed and set requirements, where no bleed water was observed on any of the grout samples after one day and all had set within 3 days of curing. This data indicates, for a well mixed product, bleed and set requirement are satisfied for the range of acidic feeds tested in this task. The yield stress measurements provide both an indication on the mixability of the salt solution with dry materials and an indication of how quickly the grout is starting to form structure. The inability to properly mix these two streams into a well mixed grout product will lead to a non-homogeneous mixture that will impact product quality. Product quality issues could be unmixed regions of dry material and hot spots having high concentrations of americium 241. Mixes that were more difficult to incorporate typically resulted in grouts with higher yield stresses. The mixability from these tests will provide Waste Solidification Building (WSB) an indication of which grouts will be more challenging to mix. The first yield stress measurements were statistically compared to a list of variables, specifically the batched chemicals used to make the acidic solutions. The first yield stress was also compared to the physical properties of the acidic solutions, physical and pH properties of the neutralized/pH adjusted solutions, and chemical and physical properties of the grout.

Hansen, E; Timothy Jones, T; Tommy Edwards, T; Alex Cozzi, A

2009-03-20T23:59:59.000Z

2

Independent Oversight Activity Report, Hanford Waste Treatment...  

Office of Environmental Management (EM)

of River Protection review of the High Level Waste Facility heating, ventilation, and air conditioning systems. Independent Oversight Activity Report, Hanford Waste Treatment...

3

UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048  

SciTech Connect (OSTI)

The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)

Chabeuf, Jean-Michel; Varet, Thierry [AREVA Site Value Development Business Unit, La Hague Site (France)] [AREVA Site Value Development Business Unit, La Hague Site (France)

2013-07-01T23:59:59.000Z

4

Extractive spectrophotometric determination of palladium from acidic high activity nuclear waste  

Science Journals Connector (OSTI)

A simple and rapid method for spectrophotometric determination of palladium from highly acidic and highly radioactive nuclear waste using ?-benzoin oxime (ABO) as...?5M to 4·10?4M in the organic phase. The molar ...

A. Dakshinamoorthy; R. K. Singh; R. H. Iyer

1994-01-01T23:59:59.000Z

5

High level nuclear waste  

SciTech Connect (OSTI)

The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-level wastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-level wastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

Crandall, J L

1980-01-01T23:59:59.000Z

6

Coupling High-Energy Radiography And Photon Activation Analysis (PAA) To Optimize The Characterization Of Nuclear Waste Packages  

SciTech Connect (OSTI)

Radiological characterization of nuclear waste packages is an industrial issue in order to select the best mode of storage. The alpha-activity, mainly due to the presence of actinides ({sup 235}U, {sup 238}U, {sup 239}Pu,...) inside the package, is one of the most important parameter to assess during the characterization. Photon Activation Analysis (PAA) is a non-destructive active method (NDA method) based on the photofission process and on the detection of delayed particles (neutrons and gammas). This technique is well-adapted to the characterization of large concrete waste packages. However, PAA methods often require a simulation step which is necessary to analyze experimental results and to quantify the global mass of actinides. The weak point of this approach is that characteristics of the package are often not well-known, these latter having a huge impact on the final simulation result. High-energy radiography, based on the use of a linear electron accelerator (LINAC), allows to visualize the content of the package and is also a performing way to tune simulation models and to optimize the characterization process by PAA. In this article, we present high-energy radiography results obtained for two different large concrete waste packages in the SAPHIR facility (Active Photon and Irradiation System). This facility is dedicated to PAA study and development and setup for a decade in CEA Saclay. We also discuss possibilities offered by the coupling between high-energy radiography and PAA techniques.

Carrel, F.; Agelou, M.; Gmar, M.; Laine, F.; Lamotte, T.; Lazaro, D.; Poumarede, B.; Rattoni, B. [CEA, LIST, F-91191, Gif-sur-Yvette (France)

2009-12-02T23:59:59.000Z

7

Independent Activity Report, Hanford Waste Treatment Plant -...  

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

Waste Treatment Plant - February 2011 Independent Activity Report, Hanford Waste Treatment Plant - February 2011 February 2011 Hanford Waste Treatment Plant Construction Quality...

8

High-Level Waste Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The guide provides the criteria for determining which DOE radioactive wastes are to be managed as high-level waste in accordance with DOE M 435.1-1.

1999-07-09T23:59:59.000Z

9

Proposed design requirements for high-integrity containers used to store, transport, and dispose of high-specific-activity, low-level radioactive wastes from Three Mile Island Unit II  

SciTech Connect (OSTI)

This report develops proposed design requirements for high integrity containers used to store, transport and/or dispose of high-activity, low-level radioactive wastes from Three Mile Island Unit II. The wastes considered are the dewatered resins produced by the EPICOR II waste treatment system used to clean-up the auxiliary building water. The radioactivity level of some of these EPICOR II liners is 1300 curies per container. These wastes may be disposed of in an intermediate depth burial (10 to 20 meter depth) facility. The proposed container design requirements are directed to ensure isolation of the waste and protection of the public health and safety.

Vigil, M.G.; Allen, G.C.; Pope, R.B.

1981-04-01T23:59:59.000Z

10

MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY  

SciTech Connect (OSTI)

This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

Bannochie, C; David Diprete, D; Ned Bibler, N

2008-12-31T23:59:59.000Z

11

Waste Loading Enhancements for Hanford Low-Activity Waste Glasses  

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

WASTE LOADING ENHANCEMENTS FOR HANFORD LOW-ACTIVITY WASTE GLASSES Albert A. Kruger, Glass Scientist DOE-WTP Project Office Engineering Division US Department of Energy Richland,...

12

Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain  

SciTech Connect (OSTI)

Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m{sup 2}, including the rock salt dump. The space required for the underground facilities amounts to 1.2 km{sup 2}, approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented.

Huertas, F.; Ulibarri, A. [ENRESA, Madrid (Spain)

1993-12-31T23:59:59.000Z

13

Independent Oversight Activity Report, Hanford Waste Treatment...  

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

and Tank Farm - January 2014 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant and Tank Farm - January 2014 January 2014 Hanford Waste...

14

Independent Oversight Activity Report, Hanford Waste Treatment...  

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

July 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - July 2013 July 2013 Operational Awareness of Waste Treatment and Immobilization...

15

Independent Oversight Activity Report, Hanford Waste Treatment...  

Energy Savers [EERE]

October 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - October 2013 October 2013 Observation of Waste Treatment and Immobilization...

16

Application of single ion activity coefficients to determine solvent extraction mechanism for components of high level nuclear waste  

SciTech Connect (OSTI)

The TRUEX solvent extraction process is being developed to remove and concentrate transuranic (TRU) elements from high-level and TRU radioactive wastes currently stored at US Department of Energy sites. Phosphoric acid is one of the chemical species of concern at the Hanford site where bismuth phosphate was used to recover plutonium. The mechanism of phosphoric acid extraction with TRUEX-NPH solvent at 25{degrees}C was determined by phosphoric acid distribution ratios, which were measured by using phosphoric acid radiotracer and a variety of aqueous phases containing different concentrations of nitric acid and nitrate ions. A model was developed for predicting phosphoric acid distribution ratios as a function of the thermodynamic activities of nitrate ion and hydrogen ion. The Generic TRUEX Model (GTM) was used to calculate these activities based on the Bromley method. The derived model supports CMPO and TBP extraction of a phosphoric acid-nitric acid complex and a CMPO-phosphoric acid complex in TRUEX-NPH solvent.

Nunez, L.; Vandegrift, G.F.

1995-12-31T23:59:59.000Z

17

Nuclear Waste Fund Activities Management Team | Department of Energy  

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

Waste Fund Activities Management Team Waste Fund Activities Management Team Nuclear Waste Fund Activities Management Team The Nuclear Waste Fund Activities Management Team has responsibility to: Manage the investments and expenditures of the Nuclear Waste Fund; Support correspondence regarding Nuclear Waste Policy Act issues raised by congressional, Inspector General, Government Accounting Office and Freedom of Information Act inquiries; and, Manage the annual fee adequacy assessment process. Applicable Documents Nuclear Waste Policy Act of 1982 Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste Standard Contract Amendment for New Reactors FY 2007 Total System Life Cycle Cost, Pub 2008 FY 2007 Fee Adequacy, Pub 2008 2009 Letter to Congress OCRWM Financial Statements for Annual Report for Years Ended

18

Optimizing High Level Waste Disposal  

SciTech Connect (OSTI)

If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being evaluated at Idaho National Laboratory and the facilities we’ve designed to evaluate options and support optimization.

Dirk Gombert

2005-09-01T23:59:59.000Z

19

SECONDARY WASTE MANAGEMENT STRATEGY FOR EARLY LOW ACTIVITY WASTE TREATMENT  

SciTech Connect (OSTI)

This study evaluates parameters relevant to River Protection Project secondary waste streams generated during Early Low Activity Waste operations and recommends a strategy for secondary waste management that considers groundwater impact, cost, and programmatic risk. The recommended strategy for managing River Protection Project secondary waste is focused on improvements in the Effiuent Treatment Facility. Baseline plans to build a Solidification Treatment Unit adjacent to Effluent Treatment Facility should be enhanced to improve solid waste performance and mitigate corrosion of tanks and piping supporting the Effiuent Treatment Facility evaporator. This approach provides a life-cycle benefit to solid waste performance and reduction of groundwater contaminants.

CRAWFORD TW

2008-07-17T23:59:59.000Z

20

Independent Oversight Activity Report, Hanford Waste Treatment...  

Office of Environmental Management (EM)

Treatment and Immobilization Plant Low Activity Waste Facility Heating, Ventilation, and Air Conditioning Systems Hazards Analysis Activities HIAR-WTP-2014-01-27 This...

Note: This page contains sample records for the topic "high activity waste" 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

Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility  

SciTech Connect (OSTI)

As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, and lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the ‘add a source’ basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.

Cherkas, Dmytro

2011-10-01T23:59:59.000Z

22

Iron Phosphate Glasses for Vitrifying DOE High Priority Nuclear Wastes  

SciTech Connect (OSTI)

Iron phosphate glasses have been studied as an alternative glass for vitrifying Department of Energy (DOE) high priority wastes. The high priority wastes were the Low Activity Waste (LAW) and the High Level Waste (HLW) with high chrome content stored at Hanford, WA, and the Sodium Bearing Waste (SBW) stored at the Idaho National Engineering and Environmental Laboratory. These wastes were recommended by Tanks Focus Area since they were expected to require special attention when vitrified in borosilicate glasses. All three of these wastes have been successfully vitrified in iron phosphate glasses at waste loadings ranging from a low of 32 wt% for the high sulfate LAW to 40 wt% for the SBW to a high of 75 wt% for the high chrome HLW. In addition to these desirable high waste loadings, the iron phosphate glasses were easily melted, typically between 950 and 1200 C, in less than 4 hours in commercial refractory oxide containers. It is noteworthy that the chemical durability of both glassy and deliberately crystallized iron phosphate wasteforms not only met, but significantly exceeded, all current DOE chemical durability requirements as measured by the Product Consistency Test (PCT) and Vapor Hydration Test (VHT). The high waste loading, low melting temperature, rapid furnace throughput (short melting time) and their outstanding chemical durability could significantly accelerate the clean up effort and reduce the time and cost of vitrifying these high priority wastes.

Kim, C.W.; Day, D.E.

2004-03-29T23:59:59.000Z

23

High Level Waste System Plan Revision 9  

SciTech Connect (OSTI)

Revision 9 of the High Level Waste System Plan documents the current operating strategy of the HLW System at SRS to receive, store, treat, and dispose of high-level waste.

Davis, N.R.; Wells, M.N.; Choi, A.S.; Paul, P.; Wise, F.E.

1998-04-01T23:59:59.000Z

24

High Level Waste Corporate Board Charter | Department of Energy  

Energy Savers [EERE]

High Level Waste Corporate Board Charter High Level Waste Corporate Board Charter High Level Waste Corporate Board Charter More Documents & Publications Corporate Board By-Laws...

25

High-Level Waste Corporate Board Presentation Archive | Department...  

Office of Environmental Management (EM)

High-Level Waste Corporate Board Presentation Archive High-Level Waste Corporate Board Presentation Archive Archived Documents High-Level Waste Corporate Board, Dr. Ins Triay...

26

Independent Oversight Activity Report, Hanford Waste Treatment...  

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

March 31 - April 10, 2014 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - March 31 - April 10, 2014 March 31 - April 10, 2014 Observation...

27

Hanford's Simulated Low Activity Waste Cast Stone Processing  

SciTech Connect (OSTI)

Cast Stone is undergoing evaluation as the supplemental treatment technology for Hanford’s (Washington) high activity waste (HAW) and low activity waste (LAW). This report will only cover the LAW Cast Stone. The programs used for this simulated Cast Stone were gradient density change, compressive strength, and salt waste form phase identification. Gradient density changes show a favorable outcome by showing uniformity even though it was hypothesized differently. Compressive strength exceeded the minimum strength required by Hanford and greater compressive strength increase seen between the uses of different salt solution The salt waste form phase is still an ongoing process as this time and could not be concluded.

Kim, Young

2013-08-20T23:59:59.000Z

28

Hanford Tank Waste - Near Source Treatment of Low Activity Waste  

SciTech Connect (OSTI)

Treatment and disposition of Hanford Site waste as currently planned consists of I 00+ waste retrievals, waste delivery through up to 8+ miles of dedicated, in-ground piping, centralized mixing and blending operations- all leading to pre-treatment combination and separation processes followed by vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The sequential nature of Tank Farm and WTP operations requires nominally 15-20 years of continuous operations before all waste can be retrieved from many Single Shell Tanks (SSTs). Also, the infrastructure necessary to mobilize and deliver the waste requires significant investment beyond that required for the WTP. Treating waste as closely as possible to individual tanks or groups- as allowed by the waste characteristics- is being investigated to determine the potential to 1) defer, reduce, and/or eliminate infrastructure requirements, and 2) significantly mitigate project risk by reducing the potential and impact of single point failures. The inventory of Hanford waste slated for processing and disposition as LAW is currently managed as high-level waste (HLW), i.e., the separation of fission products and other radionuclides has not commenced. A significant inventory ofthis waste (over 20M gallons) is in the form of precipitated saltcake maintained in single shell tanks, many of which are identified as potential leaking tanks. Retrieval and transport (as a liquid) must be staged within the waste feed delivery capability established by site infrastructure and WTP. Near Source treatment, if employed, would provide for the separation and stabilization processing necessary for waste located in remote farms (wherein most ofthe leaking tanks reside) significantly earlier than currently projected. Near Source treatment is intended to address the currently accepted site risk and also provides means to mitigate future issues likely to be faced over the coming decades. This paper describes the potential near source treatment and waste disposition options as well as the impact these options could have on reducing infrastructure requirements, project cost and mission schedule.

Ramsey, William Gene

2013-08-15T23:59:59.000Z

29

Independent Oversight Activity Report, Hanford Waste Treatment...  

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

observing a limited portion of the start of the hazard analysis (HA) for WTP Low Activity Waste (LAW) Primary Off-gas System. The primary purpose of this HSS field activity was to...

30

High Level Waste Corporate Board Charter  

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

on 24 July 2008 1 on 24 July 2008 1 Office of Environmental Management High-Level Waste Corporate Board Charter Purpose This Charter establishes the High- Level Waste (HLW) Corporate Board, (hereinafter referred to as the 'Board') within the Office of Environmental Management (EM). The Board will serve as a consensus building body to integrate the Department of Energy (DOE) HLW management and disposition activities across the EM program and, with the coordination and cooperation of other program offices, across the DOE complex. The Board will identify the need for and develop policies, planning, standards and guidance and provide the integration necessary to implement an effective and efficient national HLW program. The Board will also evaluate the implications of HLW issues and their

31

Independent Activity Report, Waste Treatment and Immobilization Plant- March 2013  

Broader source: Energy.gov [DOE]

Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review [HIAR-WTP-2013-03-18

32

An Overview of Thermoelectric Waste Heat Recovery Activities...  

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

An Overview of Thermoelectric Waste Heat Recovery Activities in Europe An Overview of Thermoelectric Waste Heat Recovery Activities in Europe An overview presentation of R&D...

33

Nuclear waste isolation activities report  

SciTech Connect (OSTI)

Included are: a report from the Deputy Assistant Secretary, a summary of recent events, new literature, a list of upcoming waste management meetings, and background information on DOE`s radwaste management programs. (DLC)

None

1980-12-01T23:59:59.000Z

34

Independent Activity Report, Hanford Waste Treatment Plant - February 2011  

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

Activity Report, Hanford Waste Treatment Plant - Activity Report, Hanford Waste Treatment Plant - February 2011 Independent Activity Report, Hanford Waste Treatment Plant - February 2011 February 2011 Hanford Waste Treatment Plant Construction Quality Assurance Review [ARPT-WTP-2011-002] The purpose of the visit was to perform a review of construction quality assurance at the Waste Treatment Plant (WTP) site activities concurrently with the Department of Energy (DOE) WTP staff. One focus area for this visit was piping and pipe support installations. Independent Activity Report, Hanford Waste Treatment Plant - February 2011 More Documents & Publications Independent Oversight Review, Waste Treatment and Immobilization Plant - August 2011 Independent Oversight Review, Waste Treatment and Immobilization Plant -

35

Northeast High-Level Radioactive Waste Transportation Task Force...  

Office of Environmental Management (EM)

Northeast High-Level Radioactive Waste Transportation Task Force Agenda Northeast High-Level Radioactive Waste Transportation Task Force Agenda Northeast High-Level Radioactive...

36

PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)  

SciTech Connect (OSTI)

The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

CERTA, P.J.

2006-02-22T23:59:59.000Z

37

Low and high Temperature Dual Thermoelectric Generation Waste...  

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

Low and high Temperature Dual Thermoelectric Generation Waste Heat Recovery System for Light-Duty Vehicles Low and high Temperature Dual Thermoelectric Generation Waste Heat...

38

Chemistry related to isolation of high-level nuclear waste  

Science Journals Connector (OSTI)

Chemistry related to isolation of high-level nuclear waste ... This article discusses the isolation of high-level nuclear waste. ... Radioactivity, Radiation, and the Chemistry of Nuclear Waste ...

Darleane C. Hoffman; Gregory R. Choppin

1986-01-01T23:59:59.000Z

39

Hanford immobilized low-activity tank waste performance assessment  

SciTech Connect (OSTI)

The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis to set requirements on the waste form and the facility design that will protect the long-term public health and safety and protect the environment.

Mann, F.M.

1998-03-26T23:59:59.000Z

40

E-Print Network 3.0 - activity radioactive waste Sample Search...  

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

waste Search Powered by Explorit Topic List Advanced Search Sample search results for: activity radioactive waste...

Note: This page contains sample records for the topic "high activity waste" 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

RCRA Notification of Regulated Waste Activity (EPA Form 8700...  

Open Energy Info (EERE)

Notification of Regulated Waste Activity (EPA Form 8700-12) Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: RCRA Notification of Regulated Waste Activity...

42

Notification of Regulated Waste Activity | Open Energy Information  

Open Energy Info (EERE)

Regulated Waste Activity Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- OtherOther: Notification of Regulated Waste ActivityLegal Published NA...

43

Iodine Solubility in Low-Activity Waste Borosilicate Glass at...  

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

Iodine Solubility in Low-Activity Waste Borosilicate Glass at 1000 °C. Iodine Solubility in Low-Activity Waste Borosilicate Glass at 1000 °C. Abstract: The purpose of this...

44

Independent Oversight Activity Report, Hanford Waste Treatment and  

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

Hanford Waste Treatment and Hanford Waste Treatment and Immobilization Plant - June 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - June 2013 June 2013 Hanford Waste Treatment and Immobilization Plant Low Activity Waste Melter Off-gas Process System Hazards Analysis Activity Observation [HIAR-WTP-2013-05-13] This Independent Activity Report documents an oversight activity conducted by the Office of Health, Safety and Security's (HSS) Office of Safety and Emergency Management Evaluations from May 13 - June 28, 2013, at the Hanford Waste Treatment and Immobilization Plant (WTP). The activity consisted of HSS staff observing a limited portion of the start of the hazard analysis (HA) for WTP Low Activity Waste (LAW) Primary Off-gas System. The primary purpose of this HSS field activity was to observe and

45

Independent Oversight Activity Report, Hanford Waste Treatment and  

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

Waste Treatment and Waste Treatment and Immobilization Plant - July 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - July 2013 July 2013 Operational Awareness of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity [HIAR-WTP-2013-07-31] This Independent Activity Report documents an oversight activity conducted by the Office of Health, Safety and Security's (HSS) Office of Safety and Emergency Management Evaluations from July 31 - August 5, 2013, at the Hanford Waste Treatment and Immobilization Plant (WTP). The activity consisted of HSS staff observing a limited portion of the hazards analysis (HA) for WTP Low Activity Waste (LAW) Melter Process system. The primary purpose of this HSS field activity was to observe and

46

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

SciTech Connect (OSTI)

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

47

High-level radioactive wastes. Supplement 1  

SciTech Connect (OSTI)

This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

McLaren, L.H. (ed.)

1984-09-01T23:59:59.000Z

48

Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle  

SciTech Connect (OSTI)

The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble components are mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and will not be available until the WTP begins operation, causing uncertainty in its composition, particularly the radionuclide content. This plan will provide an estimate of the likely composition and the basis for it, assess likely treatment technologies, identify potential disposition paths, establish target treatment limits, and recommend the testing needed to show feasibility. Two primary disposition options are proposed for investigation, one is concentration for storage in the tank farms, and the other is treatment prior to disposition in the Effluent Treatment Facility. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Recycle stream is Technetium-99 ({sup 99}Tc), a long-lived radionuclide with a half-life of 210,000 years. Technetium will not be removed from the aqueous waste in the Hanford Waste Treatment and Immobilization Plant (WTP), and will primarily end up immobilized in the LAW glass, which will be disposed in the Integrated Disposal Facility (IDF). Because {sup 99}Tc has a very long half-life and is highly mobile, it is the largest dose contributor to the Performance Assessment (PA) of the IDF. Other radionuclides that are also expected to be in appreciable concentration in the LAW Recycle are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. Although the baseline process will recycle this stream to the Pretreatment Facility, if the LAW facility begins operation first, this stream will not have a disposition path internal to WTP. One potential solution is to return the stream to the tank farms where it can be evaporated in the 242-A evaporator, or perhaps deploy an auxiliary evaporator to concentrate it prior to return to the tank farms. In either case, testing is needed to evaluat

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

2013-08-29T23:59:59.000Z

49

Regulation of geological disposal of high-level radioactive waste  

SciTech Connect (OSTI)

The Nuclear Regulatory Commission has been actively developing needed regulations over the last two years for the geological disposal of high-level radioactive waste. Technical criteria are about to be published in the form of a proposed regulation. The waste packages, underground facility, and geologic setting form the major elements of any geologic repository and the basis of a multibarrier system. Performance objectives and supporting technical criteria have been developed for each of these repository elements to provide benchmarks for scientists and engineers working in each of these major areas. 9 refs.

White, L.A.

1981-11-01T23:59:59.000Z

50

Life Extension of Aging High-Level Waste Tanks  

SciTech Connect (OSTI)

The Double Shell Tanks (DSTs) play a critical role in the Hanford High-Level Waste Treatment Complex, and therefore activities are underway to protect and better understand these tanks. The DST Life Extension Program is focused on both tank life extension and on evaluation of tank integrity. Tank life extension activities focus on understanding tank failure modes and have produced key chemistry and operations controls to minimize tank corrosion and extend useful tank life. Tank integrity program activities have developed and applied key technologies to evaluate the condition of the tank structure and predict useful tank life. Program results to date indicate that DST useful life can be extended well beyond the original design life and allow the existing tanks to fill a critical function within the Hanford High-Level Waste Treatment Complex. In addition the tank life may now be more reliably predicted, facilitating improved planning for the use and possible future replacement of these tanks.

Bryson, D.; Callahan, V.; Ostrom, M.; Bryan, W.; Berman, H.

2002-02-26T23:59:59.000Z

51

Independent Activity Report, Waste Treatment and Immobilization Plant -  

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

Waste Treatment and Immobilization Waste Treatment and Immobilization Plant - March 2013 Independent Activity Report, Waste Treatment and Immobilization Plant - March 2013 March 2013 Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review [HIAR-WTP-2013-03-18] The Office of Health, Safety and Security (HSS) staff observed a limited portion of the restart of the Hazard Analysis (HA) for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter Process (LMP) System. The primary purpose of this HSS field activity, on March 18-21, 2013, was to observe and understand the revised approach implemented by Bechtel National, Inc. (BNI), the contractor responsible for the design and construction of WTP for the U.S. Department of Energy (DOE) Office of

52

Technical basis for classification of low-activity waste fraction from Hanford site tanks  

SciTech Connect (OSTI)

The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level waste, for disposal is a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

Petersen, C.A.

1996-09-20T23:59:59.000Z

53

Permitting plan for the high-level waste interim storage  

SciTech Connect (OSTI)

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

Deffenbaugh, M.L.

1997-04-23T23:59:59.000Z

54

Independent Activity Report, Waste Isolation Pilot Plant - September...  

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

Diffusion Plant - July 2011 Independent Oversight Review, Waste Isolation Pilot Plant - April 2013 Independent Activity Report, West Valley Demonstration Project - November 2011...

55

Independent Oversight Activity Report, Hanford Waste Tank Farms...  

Office of Environmental Management (EM)

Previously Identified Items Regarding Positive Ventilation of Hanford Underground Waste Tanks HIAR-HANFORD-2013-10-28 This Independent Oversight Activity Report documents an...

56

Independent Oversight Activity Report, Hanford Waste Treatment and  

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

October 2013 October 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - October 2013 October 2013 Observation of Waste Treatment and Immobilization Plant Low Activity Waste Melter and Melter Off-gas Process System Hazards Analysis Activities [HIAR-WTP-2013-10-21] This Independent Activity Report documents an oversight activity conducted by the Office of Health, Safety and Security's (HSS) Office of Safety and Emergency Management Evaluations from October 21-31, 2013, at the Hanford Waste Treatment and Immobilization Plant (WTP). The activity consisted of HSS staff reviewing the Insight software hazard evaluation (HE) tables for hazard analysis (HA) generated to date for the WTP Low Activity Waste (LAW) Melter and Off-gas systems, observed a limited portion of the HA for the

57

Investigation of EPS Characteristics and their Effects on Waste Activated Sludge Digestion  

E-Print Network [OSTI]

Investigation of EPS Characteristics and their Effects on Waste Activated Sludge Digestion Thomas on digestibility, especially in regards to the composition of extracellular polymeric substances (EPS). Samples possibilities. Characterizations made from extraction data showed curiously high EPS disruption in Amherst

Mountziaris, T. J.

58

Stabilization of high and low solids Consolidated Incinerator Facility (CIF) waste with super cement  

SciTech Connect (OSTI)

This report details solidification activities using selected Mixed Waste Focus Area technologies with the High and Low Solid waste streams. Ceramicrete and Super Cement technologies were chosen as the best possible replacement solidification candidates for the waste streams generated by the SRS incinerator from a list of several suggested Mixed Waste Focus Area technologies. These technologies were tested, evaluated, and compared to the current Portland cement technology being employed. Recommendation of a technology for replacement depends on waste form performance, process flexibility, process complexity, and cost of equipment and/or raw materials.

Walker, B.W.

2000-01-11T23:59:59.000Z

59

West Valley Demonstration Project High-Level Waste Management  

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

DRAFT_19507_1 DRAFT_19507_1 High-Level Waste Management Bryan Bower, DOE Director - WVDP DOE High-Level Waste Corporate Board Meeting Savannah River Site April 1, 2008 West Valley Demonstration Project West Valley Demonstration Project DRAFT_19507_2 West Valley High-Level Waste To solidify the radioactive material from approximately 600,000 gallons of high-level radioactive waste into a durable, high-quality glass, both a pretreatment system to remove salts and sulfates from the waste and a vitrification system/process were designed. To solidify the radioactive material from approximately 600,000 gallons of high-level radioactive waste into a durable, high-quality glass, both a pretreatment system to remove salts and sulfates from the waste and a vitrification system/process were designed.

60

Idaho High-Level Waste & Facilities Disposition, Final Environmental...  

Office of Environmental Management (EM)

from INTEC closure activities stored in the Tank Farm, solids in the bottom of the tanks, and trace contamination from first cycle reprocessing extraction waste. SBW contains...

Note: This page contains sample records for the topic "high activity waste" 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

Followup of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process Systems Hazards Analysis Activity Review, March 2013  

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

HSS Independent Activity Report - HSS Independent Activity Report - Rev. 0 Report Number: HIAR-WTP-2013-03-18 Site: Hanford Site Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review Dates of Activity : 03/18/13 - 03/21/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff observed a limited portion of the restart of the Hazard Analysis (HA) for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter Process (LMP) System. The primary purpose of this HSS field activity, on March 18-21, 2013, was to observe and understand the revised approach

62

Followup of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process Systems Hazards Analysis Activity Review, March 2013  

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

HSS Independent Activity Report - HSS Independent Activity Report - Rev. 0 Report Number: HIAR-WTP-2013-03-18 Site: Hanford Site Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Follow-up of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity Review Dates of Activity : 03/18/13 - 03/21/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff observed a limited portion of the restart of the Hazard Analysis (HA) for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter Process (LMP) System. The primary purpose of this HSS field activity, on March 18-21, 2013, was to observe and understand the revised approach

63

DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION  

SciTech Connect (OSTI)

The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-level waste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M&O 2000a). It specifically addresses the item, ''Defense High Level Waste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to provide models and parameter values that can be used to calculate the dissolution rates for the different modes of water contact. The analyses were conducted to identify key aspects of the mechanistic model for glass dissolution to be included in the abstracted models used for PA calculations, evaluate how the models can be used to calculate bounding values of the glass dissolution rates under anticipated water contact modes in the disposal. system, and determine model parameter values for the range of potential waste glass compositions and anticipated environmental conditions. The analysis of a bounding rate also considered the effects of the buildup of glass corrosion products in the solution contacting the glass and potential effects of alteration phase formation. Note that application of the models and model parameter values is constrained to the anticipated range of HLW glass compositions and environmental conditions. The effects of processes inherent to exposure to humid air and dripping water were not modeled explicitly. Instead, the impacts of these processes on the degradation rate were taken into account by using empirically measured parameter values. These include the rates at which water sorbs onto the glass, drips onto the glass, and drips off of the glass. The dissolution rates of glasses that were exposed to humid air and dripping water measured in laboratory tests are used to estimate model parameter values for contact by humid air and dripping water in the disposal system.

W. Ebert

2001-09-20T23:59:59.000Z

64

Remote Handling Equipment for a High-Level Waste Waste Package Closure System  

SciTech Connect (OSTI)

High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

Kevin M. Croft; Scott M. Allen; Mark W. Borland

2006-04-01T23:59:59.000Z

65

Simulation of thermal stress influence on the Boom Clay kerogen (Oligocene, Belgium) in relation to long-term storage of high activity nuclear waste: I. Study of generated soluble compounds  

Science Journals Connector (OSTI)

Closed pyrolyses were performed on the Boom Clay kerogen to simulate the weak thermal stress applied during the in situ CERBERUS heating experiment (80 °C for 5 a). Two stronger thermal stresses, encompassing the range generally considered for the long-term disposal of high-activity nuclear waste (80 °C for 1 ka and 120 °C for 3 ka), were also simulated. Quantitative and qualitative studies were carried out on the products thus generated with a focus on the C12+ fraction, especially on its polar components. It thus appeared that the soluble C12+ fractions generated during these simulation experiments comprise a wide variety of polar O- and/or N-containing compounds, including carboxylic acids and phenols. The nature and/or the relative abundance of these polar compounds exhibit strong variations, with the extent of the thermal stress, reflecting the primary cracking of different types of structures with different thermal stability and the occurrence of secondary degradation reactions. These observations support the idea that the compounds, generated upon exposure of the Boom Clay kerogen to a low to moderate thermal stress, may affect the effectiveness of the geological barrier upon long-term storage of high-activity nuclear waste.

I. Deniau; S. Derenne; C. Beaucaire; H. Pitsch; C. Largeau

2005-01-01T23:59:59.000Z

66

Enterprise Assessments Operational Awareness Record, Waste Treatment...  

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

Observation of Waste Treatment and Immobilization Plant High Level Waste Facility Radioactive Liquid Waste Disposal System Hazards Analysis Activities (EA-WTP-HLW-2014-08-18(a))...

67

LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS  

SciTech Connect (OSTI)

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of {sup 99}Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility. The insoluble solids formed during these tests were primarily natrophosphate, natroxalate, and a sodium aluminosilicate compound. At the elevated temperature and 8 M [Na+], appreciable insoluble solids (1.39 wt%) were present. Cooling to room temperature and dilution of the slurry from 8 M to 5 M [Na+] resulted in a slurry containing 0.8 wt% insoluble solids. The solids (natrophosphate, natroxalate, sodium aluminum silicate, and a hydrated sodium phosphate) were relatively stable and settled quickly. Filtration rates were in the range of those observed with iron-based simulated Hanford tank sludge simulants, e.g., 6 M [Na+] Hanford tank 241-AN-102, even though their chemical speciation is considerably different. Chemical cleaning of the crossflow filter was readily accomplished with acid. As this simulant formulation was based on an average composition of a wide range of feeds using an integrated computer model, this exact composition may never be observed. But the test conditions were selected to enable comparison to the model to enable improving its chemical prediction capability.

McCabe, D.; Crawford, C.; Duignan, M.; Williams, M.; Burket, P.

2014-04-03T23:59:59.000Z

68

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING AS A SUPPLEMENTARY TREATMENT FOR HANFORD'S LOW ACTIVITY WASTE AND SECONDARY WASTES  

SciTech Connect (OSTI)

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the Savannah River National Laboratory (SRNL) to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of I-125/129 and Tc-99 to chemically resemble WTP-SW. Ninety six grams of radioactive product were made for testing. The second campaign commenced using SRS LAW chemically trimmed to look like Hanford's LAW. Six hundred grams of radioactive product were made for extensive testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

Jantzen, C.; Crawford, C.; Cozzi, A.; Bannochie, C.; Burket, P.; Daniel, G.

2011-02-24T23:59:59.000Z

69

Independent Activity Report, Waste Isolation Pilot Plant - September 2011 |  

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

Waste Isolation Pilot Plant - Waste Isolation Pilot Plant - September 2011 Independent Activity Report, Waste Isolation Pilot Plant - September 2011 September 2011 Orientation Visit to the Waste Isolation Pilot Plant [HIAR-WIPP-2011-09-07] The U.S. Department of Energy (DOE) Office of Enforcement and Oversight, within the Office of Health, Safety and Security (HSS), conducted an orientation visit to the DOE Carlsbad Field Office (CBFO) and the nuclear facility at the Waste Isolation Pilot Plant (WIPP) at Carlsbad, NM, on September 7, 2011. The purpose of the visit was to discuss the nuclear safety oversight strategy, describe the site lead program, increase HSS personnel's operational awareness of the site's activities, and identify specific activities that HSS can perform to carry out its

70

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant – March 31 – April 10, 2014  

Broader source: Energy.gov [DOE]

Observation of the Hanford Waste Treatment and Immobilization Plant Low Activity Waste Facility Hazards Analysis Activities [IAR-WTP-2014-03-31

71

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant- June 2013  

Broader source: Energy.gov [DOE]

Hanford Waste Treatment and Immobilization Plant Low Activity Waste Melter Off-gas Process System Hazards Analysis Activity Observation [HIAR-WTP-2013-05-13

72

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant – October 2013  

Broader source: Energy.gov [DOE]

Observation of Waste Treatment and Immobilization Plant Low Activity Waste Melter and Melter Off-gas Process System Hazards Analysis Activities [HIAR-WTP-2013-10-21

73

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant – February 2014  

Broader source: Energy.gov [DOE]

Hanford Waste Treatment and Immobilization Plant Low Activity Waste Facility Off-gas Systems Hazards Analysis Activities [HIAR-WTP-2014-01-27

74

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant – July 2013  

Broader source: Energy.gov [DOE]

Operational Awareness of Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity [HIAR-WTP-2013-07-31

75

Production of Biofuels from High-Acid-Value Waste Oils  

Science Journals Connector (OSTI)

Production of Biofuels from High-Acid-Value Waste Oils ... (1) Biofuel is derived from a renewable, domestic resource, thereby relieving reliance on petroleum fuel imports. ...

Junming Xu; Guomin Xiao; Yonghong Zhou; Jianchun Jiang

2011-08-27T23:59:59.000Z

76

2008 DOE Spent Nuclear Fuel and High Level Waste Inventory  

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

77

Radioactive Demonstrations Of Fluidized Bed Steam Reforming (FBSR) With Hanford Low Activity Wastes  

SciTech Connect (OSTI)

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) are being evaluated. One immobilization technology being considered is Fluidized Bed Steam Reforming (FBSR) which offers a low temperature (700-750?C) continuous method by which wastes high in organics, nitrates, sulfates/sulfides, or other aqueous components may be processed into a crystalline ceramic (mineral) waste form. The granular waste form produced by co-processing the waste with kaolin clay has been shown to be as durable as LAW glass. The FBSR granular product will be monolithed into a final waste form. The granular component is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals such as sodalite. Production of the FBSR mineral product has been demonstrated both at the industrial, engineering, pilot, and laboratory scales on simulants. Radioactive testing at SRNL commenced in late 2010 to demonstrate the technology on radioactive LAW streams which is the focus of this study.

Jantzen, C. M.; Crawford, C. L.; Burket, P. R.; Bannochie, C. J.; Daniel, W. G.; Nash, C. A.; Cozzi, A. D.; Herman, C. C.

2012-10-22T23:59:59.000Z

78

High-Level Liquid Waste Tank Integrity Workshop - 2008  

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

Liquid Waste Tank Integrity Liquid Waste Tank Integrity Workshop - 2008 Karthik Subramanian Bruce Wiersma November 2008 High Level Waste Corporate Board Meeting karthik.subramanian@srnl.doe.gov bruce.wiersma@srnl.doe.gov 2 Acknowledgements * Bruce Wiersma (SRNL) * Kayle Boomer (Hanford) * Michael T. Terry (Facilitator) * SRS - Liquid Waste Organization * Hanford Tank Farms * DOE-EM 3 Background * High level radioactive waste (HLW) tanks provide critical interim confinement for waste prior to processing and permanent disposal * Maintaining structural integrity (SI) of the tanks is a critical component of operations 4 Tank Integrity Workshop - 2008 * Discuss the HLW tank integrity technology needs based upon the evolving waste processing and tank closure requirements along with its continued storage mission

79

Summary - System Planning for Low-Activity Waste Treatment at Hanford  

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

Hanford Hanford EM Project: WTP ETR Report Date: November 2008 ETR-18 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of System Planning for Low-Activity Waste Treatment at Hanford Why DOE-EM Did This Review Construction of the facilities of the Hanford site's Waste Treatment Plant (WTP) are scheduled for completion in 2017, with radioactive waste processing scheduled to begin in 2019. An estimated 23 to 35 years will then be required to complete high-level waste (HLW) vitrification. However, vitrification of low-activity waste (LAW) may extend the WTP mission duration by decades more if supplemental LAW processing beyond the capacity of the present facility is not incorporated. The purpose of this independent review was to

80

EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms  

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

EM Waste Acceptance Product EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony Kluk/Ken Picha 2 Background * Originally Waste Acceptance Preliminary Specifications were Office of Civilian Radioactive Waste Management (RW) documents and project specific: - Defense Waste Processing Facility (PE-03, July 1989) - West Valley Demonstration Project (PE-04, January 1990) * Included many of same specifications as current version of WAPS * First version of RW Waste Acceptance System Requirements Document in January 1993 (included requirements for both SNF and HLW) * EM decided to extract requirements for HLW and put into the WAPS document 3 Background (Cont'd) * Lists technical specifications for acceptance of borosilicate HLW

Note: This page contains sample records for the topic "high activity waste" 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

Independent Activity Report, Office of River Protection Waste Treatment  

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

Office of River Protection Waste Office of River Protection Waste Treatment Plant and Tank Farms - February 2013 Independent Activity Report, Office of River Protection Waste Treatment Plant and Tank Farms - February 2013 February 2013 Site Familiarization and Introduction of New Office of Safety and Emergency Management Evaluations Site Lead for the Office of River Protection Waste Treatment Plant and Tank Farms [HIAR-HANFORD-2013-02-25] The Office of Health, Safety and Security's (HSS) Office of Safety and Emergency Management Evaluations (HS-45) assigned a new Site Lead to provide continuous oversight of activities at the Office of River Protection (ORP) Waste Treatment Plant (WTP) and tank farms. To gain familiarity with the site programs and personnel, the new Site Lead made two trips to the site, which included tours of the WTP construction site

82

Independent Oversight Activity Report, Savannah River Site Waste  

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

Independent Oversight Activity Report, Savannah River Site Waste Independent Oversight Activity Report, Savannah River Site Waste Solidification Building Independent Oversight Activity Report, Savannah River Site Waste Solidification Building May 2013 Savannah River Site Waste Solidification Building Corrective Actions from the January 2013 Report on Construction Quality of Mechanical Systems Installation and Fire Protection Design [HIAR SRS-2013-5-07] Activity Description/Purpose: Review the corrective actions being implemented by the construction contractor to address Findings 1-4, 6, and 9 from a construction quality review performed by the Office of Health, Safety and Security (HSS) (Reference 1). Meet with the SRS WSB project staff and Savannah River Nuclear Solutions (SRNS) engineers to discuss the proposed corrective actions

83

DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS  

SciTech Connect (OSTI)

This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

2009-12-30T23:59:59.000Z

84

High Level Waste ManagemenfDivision ..  

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

B.2 Environmental Documentation C. Regulatory Waste Removal Schedule (Type I, II and IV Tanks) D. Process Logic Diagram E. Process Logic Interactive Matrix F. HLW Integrated...

85

CH Packaging Operations for High Wattage Waste  

SciTech Connect (OSTI)

This document provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2006-01-06T23:59:59.000Z

86

DOE-EA-0179; Waste Form Selection for Savannah River Plant High-Level Waste  

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

48326 (F.R.) 48326 (F.R.) NOTICES DEPARTMENT OF ENERGY Compliance With the National Environmental Policy Act Proposed Finding of No Significant Impact, Selection of Borosilicate Glass as the Defense Waste Processing Facility Waste Form for High -Level Radioactive Wastes Savanah River Plant, Aiken, South Carolina Thursday, July 29, 1982 *32778 AGENCY: Energy Department. ACTION: Notice. SUMMARY: The Department of Energy (DOE) has prepared an environmental assessment (DOE/EA- 0179) on the proposed selection of borosilicate glass as the Defense Waste Processing Facility (DWPF) waste form for the immobilization of the high -level radioactive wastes generated and stored at the DOE Savannah River Plant (SRP), Aiken, South Carolina. DOE recently decided to immobilize

87

Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178  

SciTech Connect (OSTI)

The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F. [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)] [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)

2013-07-01T23:59:59.000Z

88

Reportable Nuclide Criteria for ORNL Radioactive Waste Management Activities - 13005  

SciTech Connect (OSTI)

The U.S. Department of Energy's Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee generates numerous radioactive waste streams. Many of those streams contain a large number of radionuclides with an extremely broad range of concentrations. To feasibly manage the radionuclide information, ORNL developed reportable nuclide criteria to distinguish between those nuclides in a waste stream that require waste tracking versus those nuclides of such minimal activity that do not require tracking. The criteria include tracking thresholds drawn from ORNL onsite management requirements, transportation requirements, and relevant treatment and disposal facility acceptance criteria. As a management practice, ORNL maintains waste tracking on a nuclide in a specific waste stream if it exceeds any of the reportable nuclide criteria. Nuclides in a specific waste stream that screen out as non-reportable under all these criteria may be dropped from ORNL waste tracking. The benefit of these criteria is to ensure that nuclides in a waste stream with activities which meaningfully affect safety and compliance are tracked, while documenting the basis for removing certain isotopes from further consideration. (authors)

McDowell, Kip; Forrester, Tim [Oak Ridge National Laboratory, PO Box 2008 MS-6322, Oak Ridge, TN 37831 (United States)] [Oak Ridge National Laboratory, PO Box 2008 MS-6322, Oak Ridge, TN 37831 (United States); Saunders, Mark [Fairfield Services Group, PO Box 31468, KNOxville, TN 37930 (United States)] [Fairfield Services Group, PO Box 31468, KNOxville, TN 37930 (United States)

2013-07-01T23:59:59.000Z

89

SPRU Removes High-Risk Radioactive Waste  

Broader source: Energy.gov [DOE]

NISKAYUNA, N.Y. – EM’s Separations Process Research Unit (SPRU) Disposition Project completed a significant waste-treatment campaign in February that involved the solidification of approximately 9,700 gallons of contaminated sludge and 14 shipments of the waste off-site for permanent disposal.

90

High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks  

SciTech Connect (OSTI)

The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

Troyer, G.L.

1997-03-17T23:59:59.000Z

91

Waste minimization/pollution prevention study of high-priority waste streams  

SciTech Connect (OSTI)

Although waste minimization has been practiced by the Metals and Ceramics (M&C) Division in the past, the effort has not been uniform or formalized. To establish the groundwork for continuous improvement, the Division Director initiated a more formalized waste minimization and pollution prevention program. Formalization of the division`s pollution prevention efforts in fiscal year (FY) 1993 was initiated by a more concerted effort to determine the status of waste generation from division activities. The goal for this effort was to reduce or minimize the wastes identified as having the greatest impact on human health, the environment, and costs. Two broad categories of division wastes were identified as solid/liquid wastes and those relating to energy use (primarily electricity and steam). This report presents information on the nonradioactive solid and liquid wastes generated by division activities. More specifically, the information presented was generated by teams of M&C staff members empowered by the Division Director to study specific waste streams.

Ogle, R.B. [comp.

1994-03-01T23:59:59.000Z

92

Independent Oversight Activity Report, Savannah River Site Waste  

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

Savannah River Site Waste Savannah River Site Waste Solidification Building Independent Oversight Activity Report, Savannah River Site Waste Solidification Building May 2013 Savannah River Site Waste Solidification Building Corrective Actions from the January 2013 Report on Construction Quality of Mechanical Systems Installation and Fire Protection Design [HIAR SRS-2013-5-07] Activity Description/Purpose: Review the corrective actions being implemented by the construction contractor to address Findings 1-4, 6, and 9 from a construction quality review performed by the Office of Health, Safety and Security (HSS) (Reference 1). Meet with the SRS WSB project staff and Savannah River Nuclear Solutions (SRNS) engineers to discuss the proposed corrective actions discussed in Reference 2, and clarify additional reviews to be performed by

93

Phase 1 immobilized low-activity waste operational source term  

SciTech Connect (OSTI)

This report presents an engineering analysis of the Phase 1 privatization feeds to establish an operational source term for storage and disposal of immobilized low-activity waste packages at the Hanford Site. The source term information is needed to establish a preliminary estimate of the numbers of remote-handled and contact-handled waste packages. A discussion of the uncertainties and their impact on the source term and waste package distribution is also presented. It should be noted that this study is concerned with operational impacts only. Source terms used for accident scenarios would differ due to alpha and beta radiation which were not significant in this study.

Burbank, D.A.

1998-03-06T23:59:59.000Z

94

SORPTION OF URANIUM, PLUTONIUM AND NEPTUNIUM ONTO SOLIDS PRESENT IN HIGH CAUSTIC NUCLEAR WASTE STORAGE TANKS  

SciTech Connect (OSTI)

Solids such as granular activated carbon, hematite and sodium phosphates, if present as sludge components in nuclear waste storage tanks, have been found to be capable of precipitating/sorbing actinides like plutonium, neptunium and uranium from nuclear waste storage tank supernatant liqueur. Thus, the potential may exists for the accumulation of fissile materials in such nuclear waste storage tanks during lengthy nuclear waste storage and processing. To evaluate the nuclear criticality safety in a typical nuclear waste storage tank, a study was initiated to measure the affinity of granular activated carbon, hematite and anhydrous sodium phosphate to sorb plutonium, neptunium and uranium from alkaline salt solutions. Tests with simulated and actual nuclear waste solutions established the affinity of the solids for plutonium, neptunium and uranium upon contact of the solutions with each of the solids. The removal of plutonium and neptunium from the synthetic salt solution by nuclear waste storage tank solids may be due largely to the presence of the granular activated carbon and transition metal oxides in these storage tank solids or sludge. Granular activated carbon and hematite also showed measurable affinity for both plutonium and neptunium. Sodium phosphate, used here as a reference sorbent for uranium, as expected, exhibited high affinity for uranium and neptunium, but did not show any measurable affinity for plutonium.

Oji, L; Bill Wilmarth, B; David Hobbs, D

2008-05-30T23:59:59.000Z

95

Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification  

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

Hanford Low Activity Waste (LAW) Fluidized Bed Steam Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification C.M. Jantzen and E.M. Pierce November 18, 2010 2 Participating Organizations 3 Incentive and Objectives FBSR sodium-aluminosilicate (NAS) waste form has been identified as a promising supplemental treatment technology for Hanford LAW Objectives: Reduce the risk associated with implementing the FBSR NAS waste form as a supplemental treatment technology for Hanford LAW Conduct test with actual tank wastes Use the best science to fill key data gaps Linking previous and new results together 4 Outline FBSR NAS waste form processing scales FBSR NAS waste form data/key assumptions FBSR NAS key data gaps FBSR NAS testing program 5 FBSR NAS Waste Form Processing

96

RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS  

SciTech Connect (OSTI)

High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

Fox, K.

2010-09-07T23:59:59.000Z

97

Independent Oversight Activity Report, Hanford Waste Treatment and  

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

November 2013 November 2013 Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant - November 2013 December 2013 Catholic University of America Vitreous State Laboratory Tour and Discussion of Experiments Conducted in Support of Hanford Site Waste Treatment and Immobilization Plant Select Systems Design [HIAR-VSL-2013-11-18] This Independent Activity Report documents an oversight activity conducted by the Office of Health, Safety and Security's (HSS) Office of Safety and Emergency Management Evaluations on November 18, 2013, at the Catholic University of America Vitreous State Laboratory (VSL). Bechtel National, Inc. (BNI) is the contractor responsible for the design and construction of the Hanford Site Waste Treatment and Immobilization Plant (WTP) for the

98

High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power...  

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

High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power Generation High-Efficiency Quantum-Well Thermoelectrics for Waste Heat Power Generation 2005 Diesel Engine...

99

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...  

Office of Environmental Management (EM)

: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition...

100

The High-Level Radioactive Waste Act (Manitoba, Canada)  

Broader source: Energy.gov [DOE]

Manitoba bars the storage of high-level radioactive wastes from spent nuclear fuel, not intended for research purposes, that was produced at a nuclear facility or in a nuclear reactor outside the...

Note: This page contains sample records for the topic "high activity waste" 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

Analysis of alternatives for immobilized low activity waste disposal  

SciTech Connect (OSTI)

This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program.

Burbank, D.A.

1997-10-28T23:59:59.000Z

102

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING  

SciTech Connect (OSTI)

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

S.M. Frank

2011-09-01T23:59:59.000Z

103

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant  

SciTech Connect (OSTI)

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that used for routine evaluation of feed compatibility studies for the 242-A evaporator. One of the

Adamson, Duane J.; Nash, Charles A.; McCabe, Daniel J.; Crawford, Charles L.; Wilmarth, William R.

2014-01-27T23:59:59.000Z

104

Independent Oversight Activity Report, Hanford Waste Treatment...  

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

documented safety analysis and technical safety requirements for the U.S. Department of Energy Office of River Protection's review and approval. Independent Oversight Activity...

105

Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing  

SciTech Connect (OSTI)

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

Kelly, Steven E.

2013-11-11T23:59:59.000Z

106

EIS-0062: Double-Shell Tanks for Defense High Level Waste Storage, Savannah River Site, Aiken, SC  

Broader source: Energy.gov [DOE]

This EIS analyzes the impacts of the various design alternatives for the construction of fourteen 1.3 million gallon high-activity radioactive waste tanks. The EIS further evaluates the effects of these alternative designs on tank durability, on the ease of waste retrieval from such tanks, and the choice of technology and timing for long-term storage or disposal of the wastes.

107

Technical considerations for evaluating substantially complete containment of high-level waste within the waste package  

SciTech Connect (OSTI)

This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

1990-12-01T23:59:59.000Z

108

Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532  

SciTech Connect (OSTI)

Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)

Gasbarro, Christina; Bello, Job [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States)] [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States); Bryan, Samuel; Lines, Amanda; Levitskaia, Tatiana [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)] [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)

2013-07-01T23:59:59.000Z

109

High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6  

SciTech Connect (OSTI)

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

Not Available

1994-04-01T23:59:59.000Z

110

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of high-level nuclear waste  

Science Journals Connector (OSTI)

This article is a study of the thermal effects associated with the emplacement of aged radioactive high-level wastes in a geologic repository, with emphasis on the following subjects: waste characteristics, re...

J. S. Y. Wang; D. C. Mangold; C. F. Tsang

1988-04-01T23:59:59.000Z

111

Disposal Activities and the Unique Waste Streams at the Nevada National Security Site (NNSS)  

SciTech Connect (OSTI)

This slide show documents waste disposal at the Nevada National Security Site. Topics covered include: radionuclide requirements for waste disposal; approved performance assessment (PA) for depleted uranium disposal; requirements; program approval; the Waste Acceptance Review Panel (WARP); description of the Radioactive Waste Acceptance Program (RWAP); facility evaluation; recent program accomplishments, nuclear facility safety changes; higher-activity waste stream disposal; and, large volume bulk waste streams.

Arnold, P.

2012-10-31T23:59:59.000Z

112

Maintenance Plan for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment  

SciTech Connect (OSTI)

The plan for maintaining the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (PA) is described. The plan includes expected work on PA reviews and revisions, waste reports, monitoring, other operational activities, etc.

MANN, F.M.

2000-02-09T23:59:59.000Z

113

High-Temperature Components for Rankine-Cycle-Based Waste Heat...  

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

Components for Rankine-Cycle-Based Waste Heat Recovery Systems on Combustion Engines High-Temperature Components for Rankine-Cycle-Based Waste Heat Recovery Systems on...

114

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant and Tank Farm – January 2014  

Broader source: Energy.gov [DOE]

Hanford Waste Treatment and Immobilization Plant Engineering Activities and Tank Farm Operations [HIAR-HANFORD-2014-01-13

115

High-Level Waste Corporate Board Performance Assessment Subcommittee  

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

Level Level Waste Corporate Board Performance Assessment Subcommittee John E. Marra, Ph.D. Associate Laboratory Director November 6, 2008 Richland, WA DOE-EM HLW Corporate Board Meeting Background - Performance Assessment Process Performance assessments are the fundamental risk assessment tool used by the DOE to evaluate and communicate the effectiveness and long-term impact of waste management and cleanup decisions. This includes demonstrations of compliance, NEPA analyses, and decisions about technologies and 2 analyses, and decisions about technologies and waste forms. Background - Process Perception EM-2 'Precepts' for Improved High-Level Waste Management (HLW Corporate Board Meeting - April 2008) Improved Performance Assessments (PA) The PA process is not consistently applied amongst the 3 The PA process is not consistently applied amongst the major HLW sites PA

116

Enclosure 3 DOE Response to EPA Question Regarding "High-Level Liquid Radioactive Waste"  

E-Print Network [OSTI]

to date, which is from the definitions in the Nuclear Waste Policy Act: The term "high-level radioactive of waste streams as from the applicable definition of HLW in the Nuclear Waste Policy Act. 5/11/20051 #12 defining High Level Waste: For the purpose of this statement of policy, "high-level liquid radioactive

117

High Level Waste Corporate Board Newsletter - 09/11/08  

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

UPCOMING UPCOMING EVENTS: The Low-Level Waste Federal Review Group (LFRG) in Washington, DC on 16-18 September 2008. Contact Maureen O'Dell for details (MAUREEN.O'DELL@hq.doe.gov) Next High-Level Waste Corporate Board meeting will be held at DOE- RL on 6 November 2008. Meeting details will be presented here and e- mailed to those persons with an interest to participate. Topics for discussion include but are not limited to: ï‚· Results of the Tank Integrity Workshop ï‚· Strategic Initiative Briefing ï‚· Performance Assessment Guide Proposal NEWS ITEMS 3 June 2008: WASHINGTON, DC - The U.S. Department of Energy today announced submittal of a License Application to the U.S. Nuclear Regulatory Commission seeking authorization to construct America's first repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. (http://www.ocrwm.doe.gov) 8

118

High-Level Waste Corporate Board Meeting Agenda  

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

High-Level Waste Corporate Board High-Level Waste Corporate Board Meeting Agenda Loews Hotel 1065 West Peachtree St, Atlanta, Georgia November 18, 2010 Time Topic Speaker 7:30 AM Closed Session - ratify Charter Board members 8:30 AM Welcome, Introduction, 2011 focus for HLW Corp Board Shirley Olinger 8:50 AM Introduction to Tc/I in Hanford Flowsheet ï‚· Show flowsheet w/ split locations ï‚· Describe recycle of LAW concept ï‚· Discuss baseline assumptions ï‚· Describe subsequent talks using flowsheet figure Gary Smith 9:15 AM Waste Treatment & Immobilization Plant (WTP) ï‚· Tc/I split factors (w/ and w/o recycle) ï‚· Water management (w/ and w/o recycle) Albert Kruger 9:45 AM WTP Melter/Offgas Systems Decontamination Factors ï‚· Re as a stimulant for Tc ï‚· Issues that limit Tc incorporation in LAW glass

119

DOE high-level waste tank safety program. Final report  

SciTech Connect (OSTI)

The overall objective of the work was to provide LANL with support to the DOE High-Level Waste Tank Safety Program. This effort included direct support to the DOE High-Level Waste Tank Working Groups, development of a database to track all identified safety issues, development of requirements for waste tank modernization, evaluation of external comments regarding safety-related guidance/instruction developed previously, examination of current federal and state regulations associated with DOE Tank farm operations, and performance of a conduct of operations review. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided.

NONE

1998-11-01T23:59:59.000Z

120

Production of a High-Level Waste Glass from Hanford Waste Samples  

SciTech Connect (OSTI)

The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP).

Crawford, C.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Farrara, D.M.; Ha, B.C.; Bibler, N.E.

1998-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598  

SciTech Connect (OSTI)

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

Ray, J.W. [Savannah River Remediation (United States)] [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

122

Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility  

SciTech Connect (OSTI)

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form.

Ray, J. W.; Marra, S. L.; Herman, C. C.

2013-01-09T23:59:59.000Z

123

Vitrification of High-Level Alumina Nuclear Waste  

Science Journals Connector (OSTI)

Borophosphate glass compositions have been developed for the vitrification of a high alumina calcined defense waste. The effect of substituting SiO2 and P2O5 for B2O3 on the viscosity and leach resistance was mea...

J. R. Brotzman

1980-01-01T23:59:59.000Z

124

Training Activities on Radioactive Waste Management at Moscow SIA -Radon-: Experience, Practice, Theory  

SciTech Connect (OSTI)

Management of radioactive waste relates to the category of hazardous activities. Hence the requirements to the professional level of managers and personnel working in this industry are very high. Education, training and examination of managers, operators and workers are important elements of assuring safe and efficient operation of radioactive waste management sites. The International Education Training Centre (IETC) at Moscow State Unitary Enterprise Scientific and Industrial Association 'Radon' (SIA 'Radon'), in co-operation with the International Atomic Energy Agency (IAEA), has developed expertise and provided training to waste management personnel for the last 10 years. The paper summarizes the current experience of the SIA 'Radon' in the organisation and implementation of the IAEA sponsored training and others events and outlines some of strategic educational elements, which IETC will continue to pursue in the coming years. (authors)

Batyukhnova, O.G.; Arustamov, A.E.; Dmitriev, S.A.; Agrinenko, V.V. [SUE SIA -Radon-, The 7-th Rostovsky Lane 2/14, Moscow (Russian Federation); Ojovan, M.I. [Immobilisation Science Laboratory, University of Sheffield, Sir Robert Hadfield Building (United Kingdom); Drace, Z. [International Atomic Energy Agency, Vienna (Austria)

2008-07-01T23:59:59.000Z

125

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING WITH ACUTAL HANFORD LOW ACTIVITY WASTES VERIFYING FBSR AS A SUPPLEMENTARY TREATMENT  

SciTech Connect (OSTI)

The U.S. Department of Energy's Office of River Protection is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the cleanup mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA). Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. Fluidized Bed Steam Reforming (FBSR) is one of the supplementary treatments being considered. FBSR offers a moderate temperature (700-750 C) continuous method by which LAW and other secondary wastes can be processed irrespective of whether they contain organics, nitrates/nitrites, sulfates/sulfides, chlorides, fluorides, and/or radio-nuclides like I-129 and Tc-99. Radioactive testing of Savannah River LAW (Tank 50) shimmed to resemble Hanford LAW and actual Hanford LAW (SX-105 and AN-103) have produced a ceramic (mineral) waste form which is the same as the non-radioactive waste simulants tested at the engineering scale. The radioactive testing demonstrated that the FBSR process can retain the volatile radioactive components that cannot be contained at vitrification temperatures. The radioactive and nonradioactive mineral waste forms that were produced by co-processing waste with kaolin clay in an FBSR process are shown to be as durable as LAW glass.

Jantzen, C.; Crawford, C.; Burket, P.; Bannochie, C.; Daniel, G.; Nash, C.; Cozzi, A.; Herman, C.

2012-01-12T23:59:59.000Z

126

ROAD MAP FOR DEVELOPMENT OF CRYSTAL-TOLERANT HIGH LEVEL WASTE GLASSES  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. This road map guides the research and development for formulation and processing of crystaltolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will also be addressed in this road map. The planned research described in this road map is motivated by the potential for substantial economic benefits (significant reductions in glass volumes) that will be realized if the current constraints (T1% for WTP and TL for DWPF) are approached in an appropriate and technically defensible manner for defense waste and current melter designs. The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal-tolerant high-level waste (HLW) glasses targeting high waste loadings while still meeting process related limits and melter lifetime expectancies. The modeling effort will be an iterative process, where model form and a broader range of conditions, e.g., glass composition and temperature, will evolve as additional data on crystal accumulation are gathered. Model validation steps will be included to guide the development process and ensure the value of the effort (i.e., increased waste loading and waste throughput). A summary of the stages of the road map for developing the crystal-tolerant glass approach, their estimated durations, and deliverables is provided.

Fox, K.; Peeler, D.; Herman, C.

2014-05-15T23:59:59.000Z

127

Hanford Low-Activity Waste Processing: Demonstration of the Off-Gas Recycle Flowsheet - 13443  

SciTech Connect (OSTI)

Vitrification of Hanford Low-Activity Waste (LAW) is nominally the thermal conversion and incorporation of sodium salts and radionuclides into borosilicate glass. One key radionuclide present in LAW is technetium-99. Technetium-99 is a low energy, long-lived beta emitting radionuclide present in the waste feed in concentrations on the order of 1-10 ppm. The long half-life combined with a high solubility in groundwater results in technetium-99 having considerable impact on performance modeling (as potential release to the environment) of both the waste glass and associated secondary waste products. The current Hanford Tank Waste Treatment and Immobilization Plant (WTP) process flowsheet calls for the recycle of vitrification process off-gas condensates to maximize the portion of technetium ultimately immobilized in the waste glass. This is required as technetium acts as a semi-volatile specie, i.e. considerable loss of the radionuclide to the process off-gas stream can occur during the vitrification process. To test the process flowsheet assumptions, a prototypic off-gas system with recycle capability was added to a laboratory melter (on the order of 1/200 scale) and testing performed. Key test goals included determination of the process mass balance for technetium, a non-radioactive surrogate (rhenium), and other soluble species (sulfate, halides, etc.) which are concentrated by recycling off-gas condensates. The studies performed are the initial demonstrations of process recycle for this type of liquid-fed melter system. This paper describes the process recycle system, the waste feeds processed, and experimental results. Comparisons between data gathered using process recycle and previous single pass melter testing as well as mathematical modeling simulations are also provided. (authors)

Ramsey, William G.; Esparza, Brian P. [Washington River Protection Solutions, LLC, Richland, WA 99532 (United States)] [Washington River Protection Solutions, LLC, Richland, WA 99532 (United States)

2013-07-01T23:59:59.000Z

128

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

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

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...

129

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

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

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

130

Amended Record of Decision for the Idaho High-Level Waste (HLW...  

Office of Environmental Management (EM)

Record of Decision for the Idaho High-Level Waste (HLW) and Facilities Disposition Final Environmental Impact Statement Amended Record of Decision for the Idaho High-Level Waste...

131

Low and high Temperature Dual Thermoelectric Generation Waste Heat Recovery System for Light-Duty Vehicles  

Broader source: Energy.gov [DOE]

Developing a low and high temperature dual thermoelectric generation waste heat recovery system for light-duty vehicles.

132

Deep borehole disposal of high-level radioactive waste.  

SciTech Connect (OSTI)

Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

2009-07-01T23:59:59.000Z

133

Northeast High-Level Radioactive Waste Transportation Task Force Agenda  

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

Northeast High-Level Radioactive Waste Transportation Task Force Northeast High-Level Radioactive Waste Transportation Task Force Spring Meeting - May 15, 2012 Hilton Knoxville 501 West Church Avenue, Knoxville, TN 37902-2591 Agenda (Draft #1 - 4/18/12) ______________________________________________________________________________ Tuesday, May 15 - 9:00 AM - 3:30 PM / (need meeting room name) 8:00 a.m. Continental Breakfast - served in meeting room 9:00 a.m. Task Force Business Meeting - John Giarrusso, MEMA and Rich Pinney, NJDEP Co-chairs presiding  Welcome: Introductions; Agenda Review; Announcements  2012 funding  Co-Chair Election  Rules of Procedure  Membership: members & alternates appointment status  Legislative Liaisons  Staff Regional Meeting Attendance

134

Review of High Level Waste Tanks Ultrasonic Inspection Data  

SciTech Connect (OSTI)

A review of the data collected during ultrasonic inspection of the Type I high level waste tanks has been completed. The data was analyzed for relevance to the possibility of vapor space corrosion and liquid/air interface corrosion. The review of the Type I tank UT inspection data has confirmed that the vapor space general corrosion is not an unusually aggressive phenomena and correlates well with predicted corrosion rates for steel exposed to bulk solution. The corrosion rates are seen to decrease with time as expected. The review of the temperature data did not reveal any obvious correlations between high temperatures and the occurrences of leaks. The complex nature of temperature-humidity interaction, particularly with respect to vapor corrosion requires further understanding to infer any correlation. The review of the waste level data also did not reveal any obvious correlations.

Wiersma, B

2006-03-09T23:59:59.000Z

135

Vitrified municipal waste as a host form for high-level nuclear waste  

Science Journals Connector (OSTI)

Using glass as a safe and long term hosting matrix for hazardous wastes and for the immobilization of heavy metals and nuclear wastes has become an attractive method [3]. The most known glasses used as nuclear waste

N. A. El-Alaily; E. M. Abou-Hussein…

2014-01-01T23:59:59.000Z

136

Available Options for Waste Disposal [and Discussion  

Science Journals Connector (OSTI)

...vitrified high-activity waste in properly selected deep...alternatives to present projects of waste disposal, but rather as...benefits will be different. Long-term storage of either spent fuel or vitrified waste, although not an alternative...

1986-01-01T23:59:59.000Z

137

HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY  

SciTech Connect (OSTI)

The Defense Waste Processing Facility (DWPF) is preparing for vitrification of High Level Waste (HLW) Sludge Batch 4 (SB4) in early FY2007. To support this process, the Savannah River National Laboratory (SRNL) has provided a recommendation to utilize Frit 503 for vitrifying this sludge batch, based on the composition projection provided by the Liquid Waste Organization on June 22, 2006. Frit 418 was also recommended for possible use during the transition from SB3 to SB4. A critical step in the SB4 qualification process is to demonstrate the applicability of the durability models, which are used as part of the DWPF's process control strategy, to the glass system of interest via a variability study. A variability study is an experimentally-driven assessment of the predictability and acceptability of the quality of the vitrified waste product that is anticipated from the processing of a sludge batch. At the DWPF, the durability of the vitrified waste product is not directly measured. Instead, the durability is predicted using a set of models that relate the Product Consistency Test (PCT) response of a glass to the chemical composition of that glass. In addition, a glass sample is taken during the processing of that sludge batch, the sample is transmitted to SRNL, and the durability is measured to confirm acceptance. The objective of a variability study is to demonstrate that these models are applicable to the glass composition region anticipated during the processing of the sludge batch - in this case the Frit 503 - SB4 compositional region. The success of this demonstration allows the DWPF to confidently rely on the predictions of the durability/composition models as they are used in the control of the DWPF process.

Fox, K; Tommy Edwards, T; David Peeler, D; David Best, D; Irene Reamer, I; Phyllis Workman, P

2006-10-02T23:59:59.000Z

138

E-Print Network 3.0 - active nuclear wastes Sample Search Results  

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

Powered by Explorit Topic List Advanced Search Sample search results for: active nuclear wastes Page: << < 1 2 3 4 5 > >> 1 THE UNIVERSITY OF WESTERN ONTARIO RADIATION SAFETY...

139

Toward Understanding the Effect of LowActivity Waste Glass Composition...  

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

Toward Understanding the Effect of Low-Activity Waste Glass Composition on Sulfur Solubility John D. Vienna, , Dong-Sang Kim, Isabelle S. Muller, Greg F. Piepel, ...

140

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final  

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

Idaho High-Level Waste and Facilities Disposition Final Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, DOE/EIS-0287 (September 2002)

Note: This page contains sample records for the topic "high activity waste" 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.


141

EIS-0287: Idaho High-Level Waste & Facilities Disposition | Department of  

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

7: Idaho High-Level Waste & Facilities Disposition 7: Idaho High-Level Waste & Facilities Disposition EIS-0287: Idaho High-Level Waste & Facilities Disposition SUMMARY This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD January 12, 2010 EIS-0287: Amended Record of Decision Idaho High-Level Waste and Facilities Disposition January 4, 2010

142

High Level Waste Corporate Board Newsletter - 06/03/08  

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

3 June 2008 3 June 2008 UPCOMING EVENTS: Next High-Level Waste Corporate Board meeting will be held at DOE-ID on 24 July 2008. Meeting details will be presented here and e-mailed to those persons with an interest to participate. Topics for discussion include: * Strategic Planning Initiative * Technology Development / Needs Collection / Prioritization * Waste Acceptance Product Specification This meeting will include a members-only executive session OTHER NEWS DOE SELECTS WASHINGTON RIVER PROTECTION SOLUTIONS, LLC FOR TANK OPERATIONS CONTRACT AT HANFORD SITE WASHINGTON, DC - The U.S. Department of Energy (DOE) today announced that Washington River Protection Solutions (WRPS), LLC has been selected as the tank operations contractor to store, retrieve and treat Hanford tank

143

Midwestern High-Level Radioactive Waste Transportation Project  

SciTech Connect (OSTI)

For more than half a century, the Council of State Governments has served as a common ground for the states of the nation. The Council is a nonprofit, state-supported and -directed service organization that provides research and resources, identifies trends, supplies answers and creates a network for legislative, executive and judicial branch representatives. This List of Available Resources was prepared with the support of the US Department of Energy, Cooperative Agreement No. DE-FC02-89CH10402. However, any opinions, findings, conclusions, or recommendations expressed herein are those of the author(s) and do not necessarily reflect the views of DOE. The purpose of the agreement, and reports issued pursuant to it, is to identify and analyze regional issues pertaining to the transportation of high-level radioactive waste and to inform Midwestern state officials with respect to technical issues and regulatory concerns related to waste transportation.

Dantoin, T.S.

1990-12-01T23:59:59.000Z

144

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

SciTech Connect (OSTI)

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

145

EIS-0063: Waste Management Operations, Double-Shell Tanks for Defense High Level Radioactive Waste Storage, Hanford Site, Richland, Washington  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy developed this statement to evaluate the existing tank design and consider additional specific design and safety feature alternatives for the thirteen tanks being constructed for storage of defense high-level radioactive liquid waste at the Hanford Site in Richland, Washington. This statement supplements ERDA-1538, "Final Environmental Statement on Waste Management Operation."

146

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions  

E-Print Network [OSTI]

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been originating from the reprocessing of spent nuclear fuel. Experiments using simulated nuclear waste streams

Sheffield, University of

147

Defense High-Level Waste Leaching Mechanisms Program. Final report  

SciTech Connect (OSTI)

The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

Mendel, J.E. (compiler)

1984-08-01T23:59:59.000Z

148

System Planning for Low-Activity Waste at Hanford  

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

Technical Review of System Planning Technical Review of System Planning for Low-Activity Waste Treatment at Hanford November 2008 Dr. David S. Kosson, Vanderbilt University Dr. David R. Gallay, Logistics Management Institute Dr. Ian L. Pegg, The Catholic University of America Dr. Ray G. Wymer, Oak Ridge National Laboratory (ret.) Dr. Steven Krahn, U. S. Department of Energy ii ACKNOWLEDGEMENT The Review Team thanks Mr. Ben Harp, Office of River Protection (ORP), and Mr. James Honeyman, CH2M HILL, for their exceptional support during this review. Mr. Harp was the lead Department of Energy (DOE) representative responsible for organizing reviews held on-site by the Review Team. Mr. Honeyman, and his staff, provided responsive support through technical presentations, telephone conferences, and numerous reference documents.

149

Review of the Hanford Site Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity, December 2012  

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

the Hanford Site the Hanford Site Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity December 2012 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background.......................................................................................................................................... 1 3.0 Scope and Methodology... ................................................................................................................... 1

150

Review of the Hanford Site Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity, December 2012  

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

the Hanford Site the Hanford Site Waste Treatment and Immobilization Plant Low Activity Waste Melter Process System Hazards Analysis Activity December 2012 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background.......................................................................................................................................... 1 3.0 Scope and Methodology... ................................................................................................................... 1

151

Rock mechanics activities at the Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

The application of rock mechanics at nuclear waste repositories is a true multidisciplinary effort. A description and historical summary of the Waste Isolation Pilot Plant (WIPP) is presented. Rock mechanics programs at the WIPP are outlined, and the current rock mechanics modeling philosophy of the Westinghouse Waste Isolation Division is discussed.

Francke, C. [Westinghouse Electric Corp., Carlsbad, NM (United States); Saeb, S. [International Technology Corp., Albuquerque, NM (United States)

1996-12-31T23:59:59.000Z

152

BULK VITRIFICATION TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE  

SciTech Connect (OSTI)

This report is one of four reports written to provide background information regarding immobilization technologies under consideration for supplemental immobilization of Hanford's low-activity waste. This paper is intended to provide the reader with general understanding of Bulk Vitrification and how it might be applied to immobilization of Hanford's low-activity waste.

ARD KE

2011-04-11T23:59:59.000Z

153

Tank Closure and Waste Management Environmental Impact Statement...  

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

due to releases of radionuclides and chemicals from the high-level radioactive waste tanks, Fast Flux Test Facility decommissioning, and waste management activities over long...

154

3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS  

SciTech Connect (OSTI)

This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame time-of-flight data (3D image) collected with a single laser pulse, high frame rates, direct calculation of range, blur-free images without motion distortion, no need for precision scanning mechanisms, ability to combine 3D flash LIDAR with 2D cameras for 2D texture over 3D depth, and no moving parts. The major disadvantage of the 3D flash LIDAR camera is the cost of approximately $150,000, not including the software development time and repackaging of the camera for deployment in the waste tanks.

Marzolf, A.; Folsom, M.

2010-08-31T23:59:59.000Z

155

CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315  

SciTech Connect (OSTI)

In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by closure operations. Subsequent down selection was based on compressive strength and saturated hydraulic conductivity results. Fresh slurry property results were used as the first level of screening. A high range water reducing admixture and a viscosity modifying admixture were used to adjust slurry properties to achieve flowable grouts. Adiabatic calorimeter results were used as the second level screening. The third level of screening was used to design mixes that were consistent with the fill material parameters used in the F-Tank Farm Performance Assessment which was developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closures.

Langton, C.; Burns, H.; Stefanko, D.

2012-01-10T23:59:59.000Z

156

Sequential Thermo-Hydraulic Modeling of Variably Saturated Flow in High-Level Radioactive Waste Repository  

E-Print Network [OSTI]

Sequential Thermo-Hydraulic Modeling of Variably Saturated Flow in High-Level Radioactive Waste long-lived radioactive wastes must be managed in a safe way for human health and for the environment. That is the raison why the French agency for the management of radioactive waste (ANDRA) is engaged to study

Boyer, Edmond

157

West Valley Demonstration Project Prepares to Relocate High-Level Waste |  

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

West Valley Demonstration Project Prepares to Relocate High-Level West Valley Demonstration Project Prepares to Relocate High-Level Waste West Valley Demonstration Project Prepares to Relocate High-Level Waste December 24, 2013 - 12:00pm Addthis The West Valley Demonstration Project’s high-level waste canisters will be relocated to interim onsite storage. The West Valley Demonstration Project's high-level waste canisters will be relocated to interim onsite storage. The first group of eight concrete storage casks for the West Valley Demonstration Project’s high-level waste. The first group of eight concrete storage casks for the West Valley Demonstration Project's high-level waste. Site subcontractor American DND completed demolition of the contaminated 01-14 Building in 2013. Site subcontractor American DND completed demolition of the contaminated

158

West Valley Demonstration Project Prepares to Relocate High-Level Waste |  

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

West Valley Demonstration Project Prepares to Relocate High-Level West Valley Demonstration Project Prepares to Relocate High-Level Waste West Valley Demonstration Project Prepares to Relocate High-Level Waste December 24, 2013 - 12:00pm Addthis The West Valley Demonstration Project’s high-level waste canisters will be relocated to interim onsite storage. The West Valley Demonstration Project's high-level waste canisters will be relocated to interim onsite storage. The first group of eight concrete storage casks for the West Valley Demonstration Project’s high-level waste. The first group of eight concrete storage casks for the West Valley Demonstration Project's high-level waste. Site subcontractor American DND completed demolition of the contaminated 01-14 Building in 2013. Site subcontractor American DND completed demolition of the contaminated

159

EIS-0303: Savannah River Site High-Level Waste Tank Closure | Department of  

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

03: Savannah River Site High-Level Waste Tank Closure 03: Savannah River Site High-Level Waste Tank Closure EIS-0303: Savannah River Site High-Level Waste Tank Closure SUMMARY This EIS evaluates alternatives for closing 49 high-level radioactive waste tanks and associated equipment such as evaporator systems, transfer pipelines, diversion boxes, and pump pits. DOE selected the preferred alternative identified in the Final EIS, Stabilize Tanks-Fill with Grout, to guide development and implementation of closure of the high-level waste tanks and associated equipment at the Savannah River Site. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD March 5, 2012 EIS-0303: Supplement Analysis Savannah River Site High-Level Waste Tank Closure, SC July 8, 2011 EIS-0303: Notice of Intent to Prepare an Environmental Impact Statement

160

Phosphate bonded structural products from high volume wastes  

DOE Patents [OSTI]

A method to produce structural products from benign waste is provided comprising mixing pretreated oxide with phosphoric acid to produce an acid solution, mixing the acid solution with waste particles to produce a slurry, and allowing the slurry to cure. The invention also provides for a structural material comprising waste particles enveloped by an inorganic binder. 1 fig.

Singh, D.; Wagh, A.S.

1998-12-08T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION  

SciTech Connect (OSTI)

The purpose of this calculation is to provide a dose consequence analysis of high-level waste (HLW) consisting of plutonium immobilized in vitrified HLW to be handled at the proposed Monitored Geologic Repository at Yucca Mountain for a beyond design basis event (BDBE) under expected conditions using best estimate values for each calculation parameter. In addition to the dose calculation, a plutonium respirable particle size for dose calculation use is derived. The current concept for this waste form is plutonium disks enclosed in cans immobilized in canisters of vitrified HLW (i.e., glass). The plutonium inventory at risk used for this calculation is selected from Plutonium Immobilization Project Input for Yucca Mountain Total Systems Performance Assessment (Shaw 1999). The BDBE examined in this calculation is a nonmechanistic initiating event and the sequence of events that follow to cause a radiological release. This analysis will provide the radiological releases and dose consequences for a postulated BDBE. Results may be considered in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components. This calculation uses best available technical information because the BDBE frequency is very low (i.e., less than 1.0E-6 events/year) and is not required for License Application for the Monitored Geologic Repository. The results of this calculation will not be used as part of a licensing or design basis.

J.A. Ziegler

2000-11-20T23:59:59.000Z

162

The IFR pyroprocessing for high-level waste minimization  

SciTech Connect (OSTI)

The process developed for the recycle of integral fast reactor (IFR) spent fuel utilizes a combination of pyrometallurgical and electrochemical methods and has been termed pyroprocessing. The process has been operated at full scale with simulated spent fuel using nonradioactive fission product elements. A comprehensive demonstration of the pyroprocessing of irradiated IFR fuel will begin later this year. Pyroprocessing involves the anodic dissolution of all the constituent elements of the IFR spent fuel and controlled electrotransport (electrorefining) to separate the actinide elements from the fission products present in the spent fuel. The process be applied to the processing of spent light water reactor (LWR) fuel as well, requiring only the addition of a reduction step to convert the LWR fuel as well, requiring only the addition of a reduction step to convert the LWR oxide fuel to metallic form and a separation step to separate uranium from the transuranic (TRU) elements. The TRU elements are then recovered by electroefining in the same manner as the actinides from the IFR high-level wastes arising from pyroprocessing are virtually free of actinides, and the volume of the wastes is minimized by the intrinsic characteristics of the processing of the processing method.

Laidler, J.J. (Argonne National Lab., IL (United States))

1993-01-01T23:59:59.000Z

163

Report on Abatement Activities Related to Agriculture and Waste Management  

E-Print Network [OSTI]

Guidelines for Producers 9 2.3 Best Agricultural Waste Management Plans (BAWMPs) 9 3.0 AGRICULTURAL PRACTICES COMPLAINT RESPONSE SYSTEM 8 2.1 Agricultural Waste Control Regulation and Code 9 2.2 Environmental ASSESSMENT INITIATIVE 10 3.1 Watershed Farm Practices Study 10 3.2 Ongoing Farm Practices Evaluation 12 3

164

Feasibility of lateral emplacement in very deep borehole disposal of high level nuclear waste .  

E-Print Network [OSTI]

??The U.S. Department of Energy recently filed a motion to withdraw the Nuclear Regulatory Commission license application for the High Level Waste Repository at Yucca… (more)

Gibbs, Jonathan Sutton

2010-01-01T23:59:59.000Z

165

Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement  

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

Appendix A Appendix A Site Evaluation Process A-iii DOE/EIS-0287 Idaho HLW & FD EIS TABLE OF CONTENTS Section Page Appendix A Site Evaluation Process A-1 A.1 Introduction A-1 A.2 Methodology A-1 A.3 High-Level Waste Treatment and Interim Storage Site Selection A-3 A.3.1 Identification of "Must" Criteria A-3 A.3.2 Identification of "Want" Criteria A-3 A.3.3 Identification of Candidate Sites A-3 A.3.4 Evaluation Process A-4 A.3.5 Results of Evaluation Process A-6 A.4 Low-Activity Waste Disposal Site Selection A-6 A.4.1 Identification of "Must" Criteria A-7 A.4.2 Identification of "Want" Criteria A-8 A.4.3 Identification of Candidate Sites A-8 A.4.4 Evaluation Process A-8 A.4.5 Results of Evaluation Process A-9 A.4.6 Final Selection of a Low-Activity Waste Disposal Facility

166

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

PURPOSE PURPOSE The U.S. Department of Energv SUMRY (DOE) has Dreuared this environmental impact -. . . statement (EIS) to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant (SRP) in Aiken, South Carolina. This EIS, which is both programmatic and project-specific, has been prepared in accordance with Section 102(2)(C) of the National Environmental Policy Act (NEPA) of 1969, as amended. It is intended to support broad decisions on future actions on SRP waste management activities and to provide project- related environmental input and support for project-specific decisions on pro- ceeding with cleanup activities at existing waste sites in the R- and F-Areas, establishing new waste

167

Transmutation of high-level radioactive waste - Perspectives  

E-Print Network [OSTI]

In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

2014-01-01T23:59:59.000Z

168

High Level Waste System Impacts from Acid Dissolution of Sludge  

SciTech Connect (OSTI)

This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site High Level Waste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

KETUSKY, EDWARD

2006-04-20T23:59:59.000Z

169

EIS-0120: Waste Management Activities for Groundwater Protection, Savannah River Plant, Aiken, South Carolina  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy has prepared this environmental impact statement to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant in Aiken, South Carolina.

170

EIS-0120: Waste Management Activities for Groundwater Protection, Savannah River Plant, Aiken, SC  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy has prepared this environmental impact statement to assess the environmental consequences of the implementation of modified waste management activities for hazardous, low-level radioactive, and mixed wastes for the protection of groundwater, human health, and the environment at its Savannah River Plant in Aiken, South Carolina.

171

High-Level Waste Corporate Board, Dr. Inᅢᄅs Triay  

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

Office of Environmental Management Office of Environmental Management High-Level Waste Corporate Board April 1, 2008 safety v performance v cleanup v closure M E Environmental Management Environmental Management What Are Corporate Issues? * They usually occur at multiple sites * They usually have an impact that exceeds their initial point of application. Thus, they impact: - Policies - Planning - Standards & Guidance - EM's relationship with other agencies both internal and external to DOE safety v performance v cleanup v closure M E Environmental Management Environmental Management Current Corporate Issues * Performance Assessment * Quality Assurance * Methods to Determine the Waste Inventory * Chemical Processing * Waste Forms * Actual Disposition of Waste * Waste Treatment safety v

172

RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY  

SciTech Connect (OSTI)

The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing high level waste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the high level waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

Odriscoll, R; Allan Barnes, A; Jim Coleman, J; Timothy Glover, T; Robert Hopkins, R; Dan Iverson, D; Jeff Leita, J

2008-01-15T23:59:59.000Z

173

Preliminary estimates of cost savings for defense high level waste vitrification options  

SciTech Connect (OSTI)

The potential for realizing cost savings in the disposal of defense high-level waste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-level waste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-level waste by up to $5.2 billion.

Merrill, R.A.; Chapman, C.C.

1993-09-01T23:59:59.000Z

174

Final Environmental Assessment for Waste Disposition Activities at the Paducah Site Paducah, Kentucky  

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

0-347(doc)/093002 0-347(doc)/093002 1 FINDING OF NO SIGNIFICANT IMPACT WASTE DISPOSITION ACTIVITIES AT THE PADUCAH SITE PADUCAH, KENTUCKY AGENCY: U.S. DEPARTMENT OF ENERGY ACTION: FINDING OF NO SIGNIFICANT IMPACT SUMMARY: The U.S. Department of Energy (DOE) has completed an environmental assessment (DOE/EA-1339), which is incorporated herein by reference, for proposed disposition of polychlorinated biphenyl (PCB) wastes, low-level radioactive waste (LLW), mixed low- level radioactive waste (MLLW), and transuranic (TRU) waste from the Paducah Gaseous Diffusion Plant Site (Paducah Site) in Paducah, Kentucky. All of the wastes would be transported for disposal at various locations in the United States. Based on the results of the impact analysis reported in the EA, DOE has determined that the proposed action is

175

Independent Activity Report, Office of River Protection Waste...  

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

Lead for the Office of River Protection Waste Treatment Plant and Tank Farms HIAR-HANFORD-2013-02-25 The Office of Health, Safety and Security's (HSS) Office of Safety and...

176

Independent Oversight Activity Report, Savannah River Site Waste Solidification Building  

Broader source: Energy.gov [DOE]

Savannah River Site Waste Solidification Building Corrective Actions from the January 2013 Report on Construction Quality of Mechanical Systems Installation and Fire Protection Design [HIAR SRS-2013-5-07

177

Final Tank Closure and Waste Management Environmental Impact...  

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

and treat the waste remaining in 177 underground storage tanks; store the high-level radioactive waste (HLW); dispose of the low-activity waste (LAW) at the Hanford Site...

178

STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY  

SciTech Connect (OSTI)

Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

Aaron, G.; Wilmarth, B.

2011-09-19T23:59:59.000Z

179

Road Map for Development of Crystal-Tolerant High Level Waste Glasses  

SciTech Connect (OSTI)

This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) is also addressed in this road map.

Matyas, Josef; Vienna, John D.; Peeler, David; Fox, Kevin; Herman, Connie; Kruger, Albert A.

2014-05-31T23:59:59.000Z

180

High Level Waste Management Division . H L W System Plan  

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

and Regulatory Documentation C. Regulatory Waste Removal Schedule (Type I, II and IV Tanks) D. Process Logic Diagram E. Process Logic Interactive Matrix F. HLW Integrated...

Note: This page contains sample records for the topic "high activity waste" 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.


181

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2005-04-04T23:59:59.000Z

182

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2005-04-13T23:59:59.000Z

183

Vit Plant receives and sets key air filtration equipment for Low Activity Waste Facility  

Broader source: Energy.gov [DOE]

WTP lifted a nearly 100-ton carbon bed absorber into the Low-Activity Waste Facility. This key piece of air-filtration equipment will remove mercury and acidic gases before air is channeled through...

184

WATER ACTIVITY DATA ASSESSMENT TO BE USED IN HANFORD WASTE SOLUBILITY CALCULATIONS  

SciTech Connect (OSTI)

The purpose of this report is to present and assess water activity versus ionic strength for six solutes:sodium nitrate, sodium nitrite, sodium chloride, sodium carbonate, sodium sulfate, and potassium nitrate. Water activity is given versus molality (e.g., ionic strength) and temperature. Water activity is used to estimate Hanford crystal hydrate solubility present in the waste.

DISSELKAMP RS

2011-01-06T23:59:59.000Z

185

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

SciTech Connect (OSTI)

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

186

Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks  

SciTech Connect (OSTI)

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

ROGERS, C.A.

2000-02-17T23:59:59.000Z

187

Optimisation of the Management of Higher Activity Waste in the UK - 13537  

SciTech Connect (OSTI)

The Upstream Optioneering project was created in the Nuclear Decommissioning Authority (UK) to support the development and implementation of significant opportunities to optimise activities across all the phases of the Higher Activity Waste management life cycle (i.e. retrieval, characterisation, conditioning, packaging, storage, transport and disposal). The objective of the Upstream Optioneering project is to work in conjunction with other functions within NDA and the waste producers to identify and deliver solutions to optimise the management of higher activity waste. Historically, optimisation may have occurred on aspects of the waste life cycle (considered here to include retrieval, conditioning, treatment, packaging, interim storage, transport to final end state, which may be geological disposal). By considering the waste life cycle as a whole, critical analysis of assumed constraints may lead to cost savings for the UK Tax Payer. For example, it may be possible to challenge the requirements for packaging wastes for disposal to deliver an optimised waste life cycle. It is likely that the challenges faced in the UK are shared in other countries. It is therefore likely that the opportunities identified may also apply elsewhere, with the potential for sharing information to enable value to be shared. (authors)

Walsh, Ciara; Buckley, Matthew [Nuclear Decommissioning Authority, Building 587, Curie Avenue, Harwell Oxford, Didcot, Oxfordshire, OX11 0RH (United Kingdom)] [Nuclear Decommissioning Authority, Building 587, Curie Avenue, Harwell Oxford, Didcot, Oxfordshire, OX11 0RH (United Kingdom)

2013-07-01T23:59:59.000Z

188

SWEPP PAN assay system uncertainty analysis: Active mode measurements of solidified aqueous sludge waste  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the US Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers active mode measurements of weapons grade plutonium-contaminated aqueous sludge waste contained in 208 liter drums (item description codes 1, 2, 7, 800, 803, and 807). Results of the uncertainty analysis for PAN active mode measurements of aqueous sludge indicate that a bias correction multiplier of 1.55 should be applied to the PAN aqueous sludge measurements. With the bias correction, the uncertainty bounds on the expected bias are 0 {+-} 27%. These bounds meet the Quality Assurance Program Plan requirements for radioassay systems.

Blackwood, L.G.; Harker, Y.D.; Meachum, T.R.

1997-12-01T23:59:59.000Z

189

Coupled Model for Heat and Water Transport in a High Level Waste Repository  

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

Coupled Model for Heat and Water Transport in a High Level Waste Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt This report summarizes efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a generic hypothetical high-level waste (HLW) repository in bedded salt; chemical processes of the system allow precipitation and dissolution of salt with elevated temperatures that drive water and water vapor flow around hot waste packages. Characterizing salt backfill processes is an important objective of the exercise. An evidence-based algorithm for mineral dehydration is also applied in the modeling. The Finite Element Heat and Mass transfer code (FEHM) is used to simulate coupled thermal,

190

Hanford high-level waste evaporator/crystallizer corrosion evaluation  

SciTech Connect (OSTI)

The US Department of Energy, Hanford Site nuclear reservation, located in Southeastern Washington State, is currently home to 61 Mgal of radioactive waste stored in 177 large underground storage tanks. As an intermediate waste volume reduction, the 242-A Evaporator/Crystallizer processes waste solutions from most of the operating laboratories and plants on the Hanford Site. The waste solutions are concentrated in the Evaporator/Crystallizer to a slurry of liquid and crystallized salts. This concentrated slurry is returned to Hanford Site waste tanks at a significantly reduced volume. The Washington State Department of Ecology Dangerous Waste Regulations, WAC 173-393 require that a tank system integrity assessment be completed and maintained on file at the facility for all dangerous waste tank systems. This corrosion evaluation was performed in support of the 242-A Evaporator/Crystallizer Tank System Integrity Assessment Report. This corrosion evaluation provided a comprehensive compatibility study of the component materials and corrosive environments. Materials used for the Evaporator components and piping include austenitic stainless steels (SS) (primarily ASTM A240, Type 304L) and low alloy carbon steels (CS) (primarily ASTM A53 and A106) with polymeric or asbestos gaskets at flanged connections. Building structure and secondary containment is made from ACI 301-72 Structural Concrete for Buildings and coated with a chemically resistant acrylic coating system.

Ohl, P.C.; Carlos, W.C.

1993-10-01T23:59:59.000Z

191

Chemical Environment at Waste Package Surfaces in a High-Level Radioactive Waste Repository  

SciTech Connect (OSTI)

We have conducted a series of deliquescence, boiling point, chemical transformation, and evaporation experiments to determine the composition of waters likely to contact waste package surfaces over the thermal history of the repository as it heats up and cools back down to ambient conditions. In the above-boiling period, brines will be characterized by high nitrate to chloride ratios that are stable to higher temperatures than previously predicted. This is clearly shown for the NaCl-KNO{sub 3} salt system in the deliquescence and boiling point experiments in this report. Our results show that additional thermodynamic data are needed in nitrate systems to accurately predict brine stability and composition due to salt deliquescence in dust deposited on waste package surfaces. Current YMP models capture dry-out conditions but not composition for NaCl-KNO{sub 3} brines, and they fail to predict dry-out conditions for NaCl-KNO{sub 3}-NaNO{sub 3} brines. Boiling point and deliquescence experiments are needed in NaCl-KNO{sub 3}-NaNO{sub 3} and NaCl-KNO{sub 3}-NaNO{sub 3}-Ca(NO{sub 3}){sub 2} systems to directly determine dry-out conditions and composition, because these salt mixtures are also predicted to control brine composition in the above-boiling period. Corrosion experiments are needed in high temperature and high NO{sub 3}:Cl brines to determine if nitrate inhibits corrosion in these concentrated brines at temperatures above 160 C. Chemical transformations appear to be important for pure calcium- and magnesium-chloride brines at temperatures greater than 120 C. This stems from a lack of acid gas volatility in NaCl/KNO{sub 3} based brines and by slow CO{sub 2}(g) diffusion in alkaline brines. This suggests that YMP corrosion models based on bulk solution experiments over the appropriate composition, temperature, and relative humidity range can be used to predict corrosion in thin brine films formed by salt deliquescence. In contrast to the above-boiling period, the below-boiling period is characterized predominately by NaCl based brines with minor amounts of K, NO{sub 3}, Ca, Mg, F, and Br at less than 70% relative humidity. These brines are identified as sulfate and bicarbonate brines by the chemical divide theory. Nitrate to chloride ratios are strongly tied to relative humidity and halite solubility. Once the relative humidity is low enough to produce brines saturated with respect to halite, then NO{sub 3}:Cl increases to levels and may inhibit corrosion. In addition to the more abundant NaCl-based brines some measured pore waters will evaporate towards acid NaCl-CaCl{sub 2} brines. Acid volatility also occurs with this brine type indicating that chemical transformations may be important in thin films. In contrast to the above-boiling period, comparison of our experimental data with calculated data suggest that current YMP geochemical models adequately predict in-drift chemistry in the below-boiling period.

Carroll, S; Alai, M; Craig, L; Gdowski, G; Hailey, P; Nguyen, Q A; Rard, J; Staggs, K; Sutton, M; Wolery, T

2005-05-26T23:59:59.000Z

192

Cementitious Grout for Closing SRS High Level Waste Tanks - 12315  

SciTech Connect (OSTI)

In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. Ancillary equipment abandoned in the tanks will also be filled to the extent practical. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and to be chemically reducing with a reduction potential (Eh) of -200 to -400. Grouts with this chemistry stabilize potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted to support the mass placement strategy developed by Savannah River Remediation (SRR) Closure Operations. Subsequent down selection was based on compressive strength and saturated hydraulic conductivity results. Fresh slurry property results were used as the first level of screening. A high range water reducing admixture and a viscosity modifying admixture were used to adjust slurry properties to achieve flowable grouts. Adiabatic calorimeter results were used as the second level screening. The third level of screening was used to design mixes that were consistent with the fill material parameters used in the F-Tank Farm Performance Assessment which was developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closures. The cement and slag contents of a mix selected for filling Tanks 18-F and 19-F should be limited to no more than 125 and 210 lbs/cyd, respectively, to limit the heat generated as the result of hydration reaction during curing and thereby enable mass pour placement. Trial mixes with water to total cementitious materials ratios of 0.550 to 0.580 and 125 lbs/cyd of cement and 210 lbs/cyd of slag met the strength and permeability requirements. Mix LP no.8-16 was selected for closing SRS Tanks 18-F and 19-F because it meets or exceeds the design requirements with the least amount of Portland cement and blast furnace slag. This grout is expected to flow at least 45 feet. A single point of discharge should be sufficient for unrestricted flow conditions. However, additional entry points should be identified as back-up in case restrictions in the tank impede flow. The LP no.8 series of trial mixes had surprisingly high design compressive strengths (2000 to 4000/5000 psi) which were achieved at extended curing times (28 to 90 days, respectively) given the small amount of Portland cement in the mixes (100 to 185 lbs/cyd). The grouts were flowable structural fills containing 3/8 inch gravel and concrete sand aggregate. These grouts did not segregate and require no compaction. They have low permeabilities (? 10{sup -9} cm/s) and are consequen

Langton, C.A.; Stefanko, D.B.; Burns, H.H. [Savannah River National Laboratory (United States); Waymer, J.; Mhyre, W.B. [URS Quality and Testing (United States); Herbert, J.E.; Jolly, J.C. Jr. [Savannah River Remediation, LLC, Savannah River Site, Aiken, SC 29808 (United States)

2012-07-01T23:59:59.000Z

193

Synergistic Inhibitors for Dilute High-Level Radioactive Waste  

SciTech Connect (OSTI)

Cyclic potentiodynamic polarization scans were conducted to determine the effectiveness of various combinations of anodic inhibitors in the prevention of pitting in carbon steel exposed to dilute radioactive waste. Chromate, molybdate, and phosphate were investigated as replacements for nitrite, whose effective concentrations are incompatible with the waste vitrification process. The polarization scans were performed in non-radioactive waste simulants. Their results showed that acceptable combinations of phosphate with chromate and phosphate with molybdate effectively prevented pitting corrosion. Chromate with molybdate could not replace nitrite.

Wiersma, B.J.; Zapp, P.E.

1995-11-01T23:59:59.000Z

194

High Level Waste management in Asia: R&D perspectives  

Science Journals Connector (OSTI)

The present work is an attempt to provide an overview, about the status of R&D and current trends in HLW management in Asian countries. The INIS database was selected for this purpose. Appropriate query formulations on the database, resulted in the retrieval of 4322 unique bibliographic records. Using the content analysis method, all the records were analyzed. Part One of the analysis details Scientometric R&D indicators, Part Two is a subject-based analysis, grouped under: A. Spent Fuel Recovery & Partitioning B. Waste Immobilization C. Waste Disposal and D. Waste Packaging Materials. The results of this analysis are summarized in the study.

Sangeeta Deokattey; K. Bhanumurthy; P.K. Wattal

2013-01-01T23:59:59.000Z

195

Integrated High-Level Waste System Planning - Utilizing an Integrated Systems Planning Approach to Ensure End-State Definitions are Met and Executed - 13244  

SciTech Connect (OSTI)

The Savannah River Site (SRS) is a Department of Energy site which has produced nuclear materials for national defense, research, space, and medical programs since the 1950's. As a by-product of this activity, approximately 37 million gallons of high-level liquid waste containing approximately 292 million curies of radioactivity is stored on an interim basis in 45 underground storage tanks. Originally, 51 tanks were constructed and utilized to support the mission. Four tanks have been closed and taken out of service and two are currently undergoing the closure process. The Liquid Waste System is a highly integrated operation involving safely storing liquid waste in underground storage tanks; removing, treating, and dispositioning the low-level waste fraction in grout; vitrifying the higher activity waste at the Defense Waste Processing Facility; and storing the vitrified waste in stainless steel canisters until permanent disposition. After waste removal and processing, the storage and processing facilities are decontaminated and closed. A Liquid Waste System Plan (hereinafter referred to as the Plan) was developed to integrate and document the activities required to disposition legacy and future High-Level Waste and to remove from service radioactive liquid waste tanks and facilities. It establishes and records a planning basis for waste processing in the liquid waste system through the end of the program mission. The integrated Plan which recognizes the challenges of constrained funding provides a path forward to complete the liquid waste mission within all regulatory and legal requirements. The overarching objective of the Plan is to meet all Federal Facility Agreement and Site Treatment Plan regulatory commitments on or ahead of schedule while preserving as much life cycle acceleration as possible through incorporation of numerous cost savings initiatives, elimination of non-essential scope, and deferral of other scope not on the critical path to compliance. There is currently a premium on processing and storage space in the radioactive liquid waste tank system. To enable continuation of risk reduction initiatives, the Plan establishes a processing strategy that provides tank space required to meet, or minimizes the impacts to meeting, programmatic objectives. The Plan also addresses perturbations in funding and schedule impacts. (authors)

Ling, Lawrence T. [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2205, Aiken, SC 29808 (United States)] [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2205, Aiken, SC 29808 (United States); Chew, David P. [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2426, Aiken, SC 29808 (United States)] [URS-Savannah River Remediation, Savannah River Site, Building 766-H Room 2426, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

196

Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant  

SciTech Connect (OSTI)

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. This report discusses results of preliminary radionuclide decontamination testing of the simulant. Testing examined use of Monosodium Titanate (MST) to remove {sup 90}Sr and actinides, inorganic reducing agents for {sup 99}Tc, and zeolites for {sup 137}Cs. Test results indicate that excellent removal of {sup 99}Tc was achieved using Sn(II)Cl{sub 2} as a reductant, coupled with sorption onto hydroxyapatite, even in the presence of air and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >577 in two hours. It was less effective at alkaline pH. Conversely, removal of the cesium was more effective at alka

Taylor-Pashow, Kathryn M.; Nash, Charles A.; Crawford, Charles L.; McCabe, Daniel J.; Wilmarth, William R.

2014-01-21T23:59:59.000Z

197

Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste  

SciTech Connect (OSTI)

This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

Wurm, K.J.; Miller, N.E.

1982-11-01T23:59:59.000Z

198

Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1) Introduction  

SciTech Connect (OSTI)

Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

R.A. Levich; J.S. Stuckless

2006-09-25T23:59:59.000Z

199

Glass Property Data and Models for Estimating High-Level Waste Glass Volume  

SciTech Connect (OSTI)

This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

2009-10-05T23:59:59.000Z

200

Savannah River Site (SRS) high level waste (HLW) structural integrity program  

SciTech Connect (OSTI)

The Savannah River Site has fifty-one underground tanks for radioactive waste storage and processing with doubly-contained piping systems for waste transfer. The SRS High Level Waste structural Integrity Program provides a process for evaluation and documenting material aging issues for structures, systems and components (SSC) in these facilities to maintain their confinement function. SRS has been monitoring waste, waste storage tanks, testing transfer lines and controlling waste chemistry for many years. A successful structural integrity (SI) program requires the following: detailed understanding of applicable degradation mechanisms; controlled chemistries and additions, as necessary; regular chemistry sampling and monitoring; structural capacity considerations; and a combination of on-line and periodic inspection and testing programs to provide early detection of generic degradation and verify effectiveness of the management of degradation under aging conditions identified by the SI Program. The application of these elements in the HLW SI Program achieves confinement in the facilities throughout desired service life.

Marra, J.E.; Abodishish, H.A.; Barnes, D.M.; Sindelar, R.L.; Flanders, H.E.; Houston, T.W.; Wiersma, B.J.; McNatt, F.G. Sr.; Cowfer, C.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

1995-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS  

SciTech Connect (OSTI)

The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

THIEN MG; GREER DA; TOWNSON P

2011-01-13T23:59:59.000Z

202

A COMPARISON OF HANFORD AND SAVANNAH RIVER SITE HIGH-LEVEL WASTES  

SciTech Connect (OSTI)

This study is a simple comparison of high-level waste from plutonium production stored in tanks at the Hanford and Savannah River sites. Savannah River principally used the PUREX process for plutonium separation. Hanford used the PUREX, Bismuth Phosphate, and REDOX processes, and reprocessed many wastes for recovery of uranium and fission products. Thus, Hanford has 55 distinct waste types, only 17 of which could be at Savannah River. While Hanford and Savannah River wastes both have high concentrations of sodium nitrate, caustic, iron, and aluminum, Hanford wastes have higher concentrations of several key constituents. The factors by which average concentrations are higher in Hanford salt waste than in Savannah River waste are 67 for {sup 241}Am, 4 for aluminum, 18 for chromium, 10 for fluoride, 8 for phosphate, 6 for potassium, and 2 for sulfate. The factors by which average concentrations are higher in Hanford sludges than in Savannah River sludges are 3 for chromium, 19 for fluoride, 67 for phosphate, and 6 for zirconium. Waste composition differences must be considered before a waste processing method is selected: A method may be applicable to one site but not to the other.

HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

2011-02-23T23:59:59.000Z

203

X-RAY FLUORESCENCE ANALYSIS OF HANFORD LOW ACTIVITY WASTE SIMULANTS METHOD DEVELOPMENT  

SciTech Connect (OSTI)

The x-ray fluorescence laboratory (XRF) in the Analytical Development Directorate (ADD) of the Savannah River National Laboratory (SRNL) was requested to develop an x-ray fluorescence spectrometry method for elemental characterization of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) pretreated low activity waste (LAW) stream to the LAW Vitrification Plant. The WTP is evaluating the potential for using XRF as a rapid turnaround technique to support LAW product compliance and glass former batching. The overall objective of this task was to develop an XRF analytical method that provides rapid turnaround time (<8 hours), while providing sufficient accuracy and precision to determine variations in waste.

Jurgensen, A; David Missimer, D; Ronny Rutherford, R

2007-08-08T23:59:59.000Z

204

Nuclear waste management. Quarterly progress report, October-December 1979  

SciTech Connect (OSTI)

Progress and activities are reported on the following: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization programs, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, monitoring of unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions technology, spent fuel and fuel pool integrity program, and engineered barriers. (DLC)

Platt, A.M.; Powell, J.A. (comps.)

1980-04-01T23:59:59.000Z

205

DEVELOPMENT QUALIFICATION AND DISPOSAL OF AN ALTERNATIVE IMMOBILIZED LOW-ACTIVITY WASTE FORM AT THE HANFORD SITE  

SciTech Connect (OSTI)

Demonstrating that a waste form produced by a given immobilization process is chemically and physically durable as well as compliant with disposal facility acceptance criteria is critical to the success of a waste treatment program, and must be pursued in conjunction with the maturation of the waste processing technology. Testing of waste forms produced using differing scales of processing units and classes of feeds (simulants versus actual waste) is the crux of the waste form qualification process. Testing is typically focused on leachability of constituents of concern (COCs), as well as chemical and physical durability of the waste form. A principal challenge regarding testing immobilized low-activity waste (ILAW) forms is the absence of a standard test suite or set of mandatory parameters against which waste forms may be tested, compared, and qualified for acceptance in existing and proposed nuclear waste disposal sites at Hanford and across the Department of Energy (DOE) complex. A coherent and widely applicable compliance strategy to support characterization and disposal of new waste forms is essential to enhance and accelerate the remediation of DOE tank waste. This paper provides a background summary of important entities, regulations, and considerations for nuclear waste form qualification and disposal. Against this backdrop, this paper describes a strategy for meeting and demonstrating compliance with disposal requirements emphasizing the River Protection Project (RPP) Integrated Disposal Facility (IDF) at the Hanford Site and the fluidized bed steam reforming (FBSR) mineralized low-activity waste (LAW) product stream.

SAMS TL; EDGE JA; SWANBERG DJ; ROBBINS RA

2011-01-13T23:59:59.000Z

206

Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O'Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

2001-02-01T23:59:59.000Z

207

High efficiency waste to energy facility -- Pilot plant design  

SciTech Connect (OSTI)

Waste To Energy facilities are commonly acceptable to the environment and give benefits in two main areas: one is a hygienic waste disposal and another is waste heat energy recovery to save fossil fuel consumption. Recovered energy is used for electricity supply, and it is required to increase the efficiency of refuse to electric energy conversion, and to spread the plant construction throughout the country of Japan, by the government. The national project started in 1992, and pilot plant design details were established in 1995. The objective of the project is to get 30% of energy conversion efficiency through the measure by raising the steam temperature and pressure to 500 C and 9.8 MPa respectively. The pilot plant is operating under the design conditions, which verify the success of applied technologies. This paper describes key technologies which were used to design the refuse burning boiler, which generates the highest steam temperature and pressure steam.

Orita, Norihiko; Kawahara, Yuuzou; Takahashi, Kazuyoshi; Yamauchi, Toru; Hosoda, Takuo

1998-07-01T23:59:59.000Z

208

Estimation of Failure Frequency for Type I and II High Level Waste Tanks  

SciTech Connect (OSTI)

The failure frequency of Type I and Type II High Level Waste tanks was calculated. The degradation mechanism that could lead to large break failure and the credits taken for steps taken to prevent large break failure were considered.

Subramanian, K.H.

2001-05-15T23:59:59.000Z

209

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

Broader source: Energy.gov [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...

210

Risk-informing decisions about high-level nuclear waste repositories  

E-Print Network [OSTI]

Performance assessments (PAs) are important sources of information for societal decisions in high-level radioactive waste (HLW) management, particularly in evaluating safety cases for proposed HLW repository development. ...

Ghosh, Suchandra Tina, 1973-

2004-01-01T23:59:59.000Z

211

Design of a high-level waste repository system for the United States  

E-Print Network [OSTI]

This report presents a conceptual design for a High Level Waste disposal system for fuel discharged by U.S. commercial power reactors, using the Yucca Mountain repository site recently designated by federal legislation. ...

Driscoll, Michael J.

1988-01-01T23:59:59.000Z

212

High-Performance Thermoelectric Devices Based on Abundant Silicide Materials for Vehicle Waste Heat Recovery  

Broader source: Energy.gov [DOE]

Development of high-performance thermoelectric devices for vehicle waste heat recovery will include fundamental research to use abundant promising low-cost thermoelectric materials, thermal management and interfaces design, and metrology

213

Glass Formulation and Testing for U.S. High-Level Tank Wastes?Project...  

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

Formulation and Testing for U.S. High-Level Tank Wastes-Project 17210 JD Vienna, Pacific Northwest National Laboratory, Richland, WA, USA AA Kruger, U.S. Department of Energy,...

214

Calculations of Induced Activity in the ATLAS Experiment for Nuclear Waste Zoning.  

E-Print Network [OSTI]

Extensive calculations were performed with the general activation formula using the fluxes of high-energy hadrons and low-energy neutrons previously obtained from simulations with the GCALOR code of the ATLAS detector. Three sets of proton cross-sections were used for hadrons energy above 20 MeV: (a) one set calculated with the YIELDX code (i.e., the Silberberg-Tsao formula of partial proton spallation cross-sections), (b) one set calculated with the Rudstam formula, and (c) the â??best-estimate' dataset which was a compilation of the available experimental and calculated data. In the energy region below 20 MeV, neutron activation cross-sections were taken from evaluated nuclear data files. The activity of each nuclide for a predefined operation scenario (i.e., number and duration of irradiation and shutdown cycles) was normalized to reference values taken from the European or Swiss legislations, to obtain an aggregate estimate of the radiological hazard comparable with a nuclear waste zoning definition cr...

Morev, M N

2007-01-01T23:59:59.000Z

215

New high-level waste management technology for IFR pyroprocessing wastes  

SciTech Connect (OSTI)

The pyrochemical electrorefining process for recovery of actinides in spent fuel from the Integral Fast Reactor accumulates fission product wastes as chlorides dissolved in molten LiCI-KCI and as metals, some of which are in molten cadmium. Pyrochemical processes are being developed to recover uranium and transuranium elements for return to the reactor, and to separate and immobilize fission products in suitable waste forms. Solvent cadmium is recycled within the process. Electrolyte salt is treated in a series of salt/cadmium extraction steps; it is also returned to the process. Salt-borne fission products are concentrated on a zeolite bed that is converted to a stable, leach-resistant mineral. Rare earth fission products from the salt, noble metal fission products, and cladding hulls are dispersed in a metal matrix.

Ackerman, J.P.; Johnson, T.R.

1993-09-01T23:59:59.000Z

216

HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS  

SciTech Connect (OSTI)

The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

THIEN MG; WELLS BE; ADAMSON DJ

2010-01-14T23:59:59.000Z

217

HIGH TEMPERATURE TREATMENT OF INTERMEDIATE-LEVEL RADIOACTIVE WASTES - SIA RADON EXPERIENCE  

SciTech Connect (OSTI)

This review describes high temperature methods of low- and intermediate-level radioactive waste (LILW) treatment currently used at SIA Radon. Solid and liquid organic and mixed organic and inorganic wastes are subjected to plasma heating in a shaft furnace with formation of stable leach resistant slag suitable for disposal in near-surface repositories. Liquid inorganic radioactive waste is vitrified in a cold crucible based plant with borosilicate glass productivity up to 75 kg/h. Radioactive silts from settlers are heat-treated at 500-700 0C in electric furnace forming cake following by cake crushing, charging into 200 L barrels and soaking with cement grout. Various thermochemical technologies for decontamination of metallic, asphalt, and concrete surfaces, treatment of organic wastes (spent ion-exchange resins, polymers, medical and biological wastes), batch vitrification of incinerator ashes, calcines, spent inorganic sorbents, contaminated soil, treatment of carbon containing 14C nuclide, reactor graphite, lubricants have been developed and implemented.

Sobolev, I.A.; Dmitriev, S.A.; Lifanov, F.A.; Kobelev, A.P.; Popkov, V.N.; Polkanov, M.A.; Savkin, A.E.; Varlakov, A.P.; Karlin, S.V.; Stefanovsky, S.V.; Karlina, O.K.; Semenov, K.N.

2003-02-27T23:59:59.000Z

218

Overview of Hanford Site High-Level Waste Tank Gas and Vapor Dynamics  

SciTech Connect (OSTI)

Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-level waste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

Huckaby, James L.; Mahoney, Lenna A.; Droppo, James G.; Meacham, Joseph E.

2004-08-31T23:59:59.000Z

219

High Level Waste Corporate Board Newsletter - 06/03/09  

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

UPCOMING EVENTS: UPCOMING EVENTS: Tank Waste Corporate Board Oak Ridge National Laboratory Oak Ridge, Tennessee 28 - 29 July 2009 The Board meeting will be preceded by a tour of the Radiochemical Engineering and Development Center on the afternoon of Tuesday, 28 July, and the meeting is planned for a full day on Wednesday, 29 July. Agenda Items include: ï‚· Future Directions for DOE Office of Nuclear Energy ï‚· Robotic Arm for Tank Cleaning ï‚· AREVA Mobile Hot Cell ï‚· Integrated Project Team Report ï‚· DOE Nuclear Safety Research and Development Coordinating Committee ï‚· Melton Valley Clean-Up: Lessons Learned ï‚· Chemical Cleaning of Waste Tanks at Savannah River - F Tank Farm Closure Project ï‚· Structural Integrity of Single Shell Tanks ï‚· Report from the Performance

220

Independent Oversight Activity Report, Savannah River Site Waste...  

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

Building Corrective Actions from the January 2013 Report on Construction Quality of Mechanical Systems Installation and Fire Protection Design HIAR SRS-2013-5-07 Activity...

Note: This page contains sample records for the topic "high activity waste" 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

Waste management regulatory compliance issues related to D&D activities at Oak Ridge National Laboratory (ORNL)  

SciTech Connect (OSTI)

The waste management activities at ORNL related to the decontamination and decommissioning (D&D) of radioactively contaminated buildings are divided into four categories: Operational facilities, inactive or surplus facilities, future facilities planning, and D&D activities. This paper only discusses regulatory issues related to inactive or surplus facilities. Additionally, rather than attempting to address all resulting waste streams and related regulations, this paper highlights only a few of the ORNL waste streams that present key regulatory issues.

Hitch, J.P.; Arnold, S.E.; Burwinkle, T.; Daugherty, D.

1994-09-01T23:59:59.000Z

222

Management Activities for Retrieved and Newly Generated Transuranic Wastes Savannah River Plant  

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

8 WL 253648 (F.R.) 8 WL 253648 (F.R.) NOTICES DEPARTMENT OF ENERGY Finding of No Significant Impact; Transuranic Waste Management Activities at the Savannah River Plant, Aiken, SC Tuesday, August 30, 1988 *33172 AGENCY: Department of Energy. ACTION: Finding of No Significant Impact. SUMMARY: The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA -0315, for transuranic (TRU) waste management activities at DOE's Savannah River Plant (SRP), including the construction and operation of a new TRU Waste Processing Facility. Based on analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact

223

Evaluation of high-level waste pretreatment processes with an approximate reasoning model  

SciTech Connect (OSTI)

The development of an approximate-reasoning (AR)-based model to analyze pretreatment options for high-level waste is presented. AR methods are used to emulate the processes used by experts in arriving at a judgment. In this paper, the authors first consider two specific issues in applying AR to the analysis of pretreatment options. They examine how to combine quantitative and qualitative evidence to infer the acceptability of a process result using the example of cesium content in low-level waste. They then demonstrate the use of simple physical models to structure expert elicitation and to produce inferences consistent with a problem involving waste particle size effects.

Bott, T.F.; Eisenhawer, S.W.; Agnew, S.F.

1999-04-01T23:59:59.000Z

224

Enterprise Assessments Operational Awareness Record, Waste Treatment and Immobilization Plant – December 2014  

Broader source: Energy.gov [DOE]

Operational Awareness Record for the Observation of Waste Treatment and Immobilization Plant High Level Waste Facility Radioactive Liquid Waste Disposal System Hazards Analysis Activities (EA-WTP-HLW-2014-08-18(a))

225

CAST STONE TECHNOLOGY FOR THE TREATMENT AND IMMOBILIZATION OF LOW-ACTIVITY WASTE  

SciTech Connect (OSTI)

Cast stone technology is being evaluated for potential application in the treatment and immobilization of Hanford low-activity waste. The purpose of this document is to provide background information on cast stone technology. The information provided in the report is mainly based on a pre-conceptual design completed in 2003.

MINWALL HJ

2011-04-08T23:59:59.000Z

226

Preliminary Closure Plan for the Immobilized Low Activity Waste (ILAW) Disposal Facility  

SciTech Connect (OSTI)

This document describes the preliminary plans for closure of the Immobilized Low-Activity Waste (ILAW) disposal facility to be built by the Office of River Protection at the Hanford site in southeastern Washington. The facility will provide near-surface disposal of up to 204,000 cubic meters of ILAW in engineered trenches with modified RCRA Subtitle C closure barriers.

BURBANK, D.A.

2000-08-31T23:59:59.000Z

227

Impact of EPS on Digestion of Waste Activate Sludge Thomas Gostanian  

E-Print Network [OSTI]

Impact of EPS on Digestion of Waste Activate Sludge Thomas Gostanian Faculty Mentor: Professor Chul of WAS is the make-up of the Extracellular Polymeric Substances (EPS) surrounding the cells, which acts as a glue/inorganics in influent, temperature, reactor flow types, etc). Certain EPS complexes are vulnerable to specific

Mountziaris, T. J.

228

Final Environmental Impact Statement (Supplement to ERDA-1537, September 1977) Waste Management Operations Double-Shell Tanks for Defense High-Level Radioactive Waste Storage Savannah River Plant  

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

Do Do E/EIS-0062 FINAL ENVIRONMENTAL IMPACT mATEIUIENT (Supplement to ERDA-1537, September 1977) Waste ~ Management Operations Savannah River Plant ! Aiken, South Carolina Double-Shell Tanks for Defense High-Level Radioactive Waste Storage April 1980 U.S. DEPARTMENT OF ENERGY WASHINGTON. D.C.20545 1980 WL 94273 (F.R.) NOTICES DEPARTMENT OF ENERGY Office of Deputy Assistant Secretary for Nuclear Waste Management Double-Shell Tanks for Defense High-Level Radioactive Waste Storage, Savannah River Plant, Aiken, S.C. Wednesday, July 9, 1980 *46154 Record of Decision Decision. The decision has been made to complete the construction of the 14 double-shell tanks and use them to store defense high-level radioactive waste at the Savannah River Plant (SRP). Background. The SRP, located near Aiken, South Carolina, is a major installation of the

229

Processing and utilizing high heat value, low ash alternative fuels from urban solid waste  

SciTech Connect (OSTI)

The history of technologies in the US that recover energy from urban solid waste is relatively short. Most of the technology as we know it evolved over the past 25 years. This evolution led to the development of about 100 modern mass burn and RDF type waste-to-energy plants and numerous small modular combustion systems, which collectively are handling about 20%, or about 40 million tons per year, of the nations municipal solid waste. Technologies also evolved during this period to co-fire urban waste materials with other fuels or selectively burn specific waste streams as primary fuels. A growing number of second or third generation urban waste fuels projects are being developed. This presentation discusses new direction in the power generating industry aimed at recovery and utilization of clean, high heat value, low ash alternative fuels from municipal and industrial solid waste. It reviews a spectrum of alternative fuels for feasible recovery and reuse, with new opportunities emerging for urban fuels processors providing fuels in the 6,000--15,000 BTU/LB range for off premises use.

Smith, M.L. [M.L. Smith Environmental and Associates, Tinley Park, IL (United States)

1995-10-01T23:59:59.000Z

230

Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates  

SciTech Connect (OSTI)

The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

2014-01-13T23:59:59.000Z

231

Replacement inhibitors for high level waste tank cooling coils  

SciTech Connect (OSTI)

Sodium chromate has been an effective corrosion inhibitor for the cooling coil systems in Savannah River Site (SRS) waste tanks for over 40 years. Due to age and operating history, cooling coils occasionally fail allowing chromated water to leak into the environment. The costs of reporting and cleaning up chromate spills became significant enough for SRS to consider alternate inhibitors. Confirmatory tests were performed to assess the effectiveness of three alternative corrosion inhibitor systems for the waste tank cooling water systems: (1) sodium molybdate (250 ppm as Mo)/sodium hydroxide (pH 10), (2) sodium molybdate (50 ppm as Mo)/sodium silicate (50 ppm as Si), and (3) sodium nitrite (500 ppm)/ sodium hydroxide (0.01 M). The tests were conducted under stagnant conditions to simulate a worst-case scenario. The results showed that these inhibitors were as effective as chromate at minimizing general corrosion at solution temperatures between 30--70 C. However, the initiation of localized attack in crevice regions, in solutions containing the alternative inhibitors at 70 C, was observed. Also, for the nitrite and the molybdate systems to be effective, suitable biocide is needed. On the other hand, interval coupon tests showed that the molybdate inhibitor systems prevented significant propagation of the localized attack.

Wiersma, B.J.; Hsu, T.C. [Westinghouse Savannah River Co., Aiken, SC (United States)

1996-10-01T23:59:59.000Z

232

EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes  

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

023: Long-Term Management of Defense High-Level Radioactive 023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization) Savannah River Plant, Aiken, South Carolina EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization) Savannah River Plant, Aiken, South Carolina SUMMARY This EIS analyzes the potential environmental implications of the proposed continuation of a large Federal research and development (R&D) program directed toward the immobilization of the high-level radioactive wastes resulting from chemical separations operations for defense radionuclides production at the DOE Savannah River Plant (SRP) near Aiken, South Carolina. PUBLIC COMMENT OPPORTUNITIES None available at this time.

233

LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT  

SciTech Connect (OSTI)

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate to the insoluble technetium dioxide. The reducing agents were tried with and without sorbents.

Taylor-Pashow, K.; Nash, C.; McCabe, D.

2014-09-29T23:59:59.000Z

234

Clean solid biofuel production from high moisture content waste biomass employing hydrothermal treatment  

Science Journals Connector (OSTI)

Abstract Our society currently faces three challenges, including resource depletion, waste accumulation and environmental degradation, leading to rapidly escalating raw material costs and increasingly expensive and restrictive waste disposal legislation. This work aims to produce clean solid biofuel from high moisture content waste biomass (bio-waste) with high nitrogen (N)/chlorine (Cl) content by mild hydrothermal (HT) conversion processes. The newest results are summarized and discussed in terms of the mechanical dewatering and upgrading, dechlorination, denitrification and coalification resulting from the HT pretreatment. Moreover, both the mono-combustion and co-combustion characteristics of the solid fuel are reviewed by concentrating on the pollutants emission control, especially the NO emission properties. In addition, the feasibility of this HT solid biofuel production process is also discussed in terms of “Energy Balance and economic viability”. As an alternative to dry combustion/dry pyrolysis/co-combustion, the HT process, combining the dehydration and decarboxylation of a biomass to raise its carbon content aiming to achieve a higher calorific value, opens up the field of potential feedstock for lignite-like solid biofuel production from a wide range of nontraditional renewable and plentiful wet agricultural residues, sludge and municipal wastes. It would contribute to a wider application of HT pretreatment bio-wastes for safe disposal and energy recycling.

Peitao Zhao; Yafei Shen; Shifu Ge; Zhenqian Chen; Kunio Yoshikawa

2014-01-01T23:59:59.000Z

235

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

Els-o120 Els-o120 Final Environmental Impact Statement I Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina of Energy 1 COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy ACTIVITY: Final Environmental Impact Statement, Waste Management I TC Activities for Groundwater Protection at the Savannah River Plant, Aiken, South Carolina. CONTACT: ABSTRACT: Additional information concerning this Statement can be obtained from: Mr. S. R. Wright Director, Environmental Division U.S. Department of Energy Savannah River Operations Office Post Office Box A Aiken, South Carolina 29802 (803) 725-3957 I TC For general information on the Department of Energy qs EIS process contact: Office of the Assistant Secretary for Environment, Safety, and Health U.S. Department of Energy Attn: Ms. Carol Bergstrom (EH-25) Acting Director, Office of

236

Unrestricted disposal of minimal activity levels of radioactive wastes: exposure and risk calculations  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission is currently considering revision of rule 10 CFR Part 20, which covers disposal of solid wastes containing minimal radioactivity. In support of these revised rules, we have evaluated the consequences of disposing of four waste streams at four types of disposal areas located in three different geographic regions. Consequences are expressed in terms of human exposures and associated health effects. Each geographic region has its own climate and geology. Example waste streams, waste disposal methods, and geographic regions chosen for this study are clearly specified. Monetary consequences of minimal activity waste disposal are briefly discussed. The PRESTO methodology was used to evaluate radionuclide transport and health effects. This methodology was developed to assess radiological impacts to a static local population for a 1000-year period following disposal. Pathways and processes of transit from the trench to exposed populations included the following considerations: groundwater transport, overland flow, erosion, surface water dilution, resuspension, atmospheric transport, deposition, inhalation, and ingestion of contaminated beef, milk, crops, and water. 12 references, 2 figures, 8 tables.

Fields, D.E.; Emerson, C.J.

1984-08-01T23:59:59.000Z

237

Large-dimension, high-ZT Thermoelectric Nanocomposites for High-Power High-efficiency Waste Heat Recovery for Electricity Generation  

Broader source: Energy.gov [DOE]

Large-dimension, high-ZT BiTe and Pb-based nanocomposites produced with a low-cost scalable process were used for development and testing of TE module prototypes, and demonstration of a waste heat recovery system

238

The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project  

SciTech Connect (OSTI)

The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

Presgrove, S.B. (Bechtel Savannah River, Inc., North Augusta, SC (United States))

1992-01-01T23:59:59.000Z

239

The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project  

SciTech Connect (OSTI)

The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

Presgrove, S.B. [Bechtel Savannah River, Inc., North Augusta, SC (United States)

1992-08-01T23:59:59.000Z

240

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

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

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

Note: This page contains sample records for the topic "high activity waste" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


241

DEPARTMENT OF ENERGY Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford  

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

Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford Site, Richland, Washington; Record of Decision (ROO). This Record of Decision has been prepared pursuant to the Council on Environme~tal Quality ~egulations for Implementing the Procedural Provisions of the National Environmental Pol icy Act (NEPAl (40 CFR Parts 1500-1508) and the Department of Energy NEPA Guidelines (52 FR 47662, December 15, 1987). It is based on DOE's "Environmental Impact Statement for the Oi sposal of Hanford Defense High-Level, Transuranic, and Tank Wastes'' (OOE/EIS-0113) and consideration of ~11 public and agency comments received on the Environmental Impact Statement (EIS). fJECISION The decision is to implement the ''Preferred Alternative'' as discussed in

242

Constructing a Risk Controversy: The Case of a Proposed High-Level Nuclear Waste Repository on the Skull Valley Goshute.  

E-Print Network [OSTI]

?? This thesis is a qualitative case study of a risk controversy generated by a proposal to construct a high-level nuclear waste repository on the… (more)

Jones, Taunya J.

2005-01-01T23:59:59.000Z

243

Steam reforming utilizing high activity catalyst  

SciTech Connect (OSTI)

High activity, sulfur tolerant steam reforming catalysts are described comprising rhodium or nickel supported on lanthanum stabilized alumina or magnesium promoted lanthanum stabilized alumina. The catalysts have improved activity over conventionally used catalysts in the presence of sulfur containing hydrocarbon fuel (such as No. 2 fuel oil) in a steam reforming environment. The material has particular utility in autothermal, tubular, cyclic and adiabatic steam reforming processes.

Setzer, H. J.

1985-03-05T23:59:59.000Z

244

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1995  

SciTech Connect (OSTI)

To assist the Department of Energy (DOE) in fulfilling its responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985, the National Low-Level Waste Management Program (NLLWMP) outlines the key activities that the NLLWMP will accomplish in the following fiscal year. Additional activities are added during the fiscal year as necessary to accomplish programmatic goals. This report summarizes the activities and accomplishments of the NLLWMP during fiscal year 1995.

Forman, S.

1995-12-01T23:59:59.000Z

245

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1996  

SciTech Connect (OSTI)

To assist the Department of Energy (DOE) in fulfilling its responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985, the National Low-Level Waste Management Program (NLLWMP) outlines the key activities tat the NLLWMP will accomplish in the following fiscal year. Additional activities are added during the fiscal year as necessary to accomplish programmatic goals. This report summarizes the activities and accomplishments of the NLLWMP during Fiscal Year 1996.

Garcia, R.S.

1996-12-01T23:59:59.000Z

246

PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS  

SciTech Connect (OSTI)

The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt % Pu in the glass did not adversely impact glass viscosity (as assessed using Hf surrogate) or glass durability. Finally, evaluation of DWPF glass pour samples that had Pu concentrations below the 897 g/m{sup 3} limit showed that Pu concentrations in the glass pour stream were close to targeted compositions in the melter feed indicating that Pu neither volatilized from the melt nor stratified in the melter when processed in the DWPF melter.

Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

2011-01-04T23:59:59.000Z

247

Methods of vitrifying waste with low melting high lithia glass compositions  

DOE Patents [OSTI]

The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

Jantzen, Carol M. (Aiken, SC); Pickett, John B. (Aiken, SC); Cicero-Herman, Connie A. (Aiken, SC); Marra, James C. (Aiken, SC)

2001-01-01T23:59:59.000Z

248

Reference design and operations for deep borehole disposal of high-level radioactive waste.  

SciTech Connect (OSTI)

A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall, the results of the reference design development and the cost analysis support the technical feasibility of the deep borehole disposal concept for high-level radioactive waste.

Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

2011-10-01T23:59:59.000Z

249

STRONTIUM AND ACTINIDE SEPARATIONS FROM HIGH LEVEL NUCLEAR WASTE SOLUTIONS USING MONOSODIUM TITANATE 1. SIMULANT TESTING  

SciTech Connect (OSTI)

High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove {sup 137}Cs, {sup 90}Sr and alpha-emitting radionuclides (i.e., actinides) prior to disposal. Separation processes planned at SRS include caustic side solvent extraction, for {sup 137}Cs removal, and ion exchange/sorption of {sup 90}Sr and alpha-emitting radionuclides with an inorganic material, monosodium titanate (MST). The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes {sup 238}Pu, {sup 239}Pu and {sup 240}Pu. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from simulated waste solutions. These tests evaluated the influence of ionic strength, temperature, solution composition and the oxidation state of plutonium.

HOBBS, D. T.; BARNES, M. J.; PULMANO, R. L.; MARSHALL, K. M.; EDWARDS, T. B.; BRONIKOWSKI, M. G.; FINK, S. D.

2005-04-14T23:59:59.000Z

250

Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)  

SciTech Connect (OSTI)

Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

Not Available

1980-10-01T23:59:59.000Z

251

Review of the Waste Isolation Pilot Plant Work Planning and Control Activities, April 2013  

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

Independent Oversight Review of the Independent Oversight Review of the Waste Isolation Pilot Plant Work Planning and Control Activities April 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Scope.................................................................................................................................................... 1 3.0 Background........................................................................................................................................... 1 4.0 Methodology......................................................................................................................................... 2

252

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2  

SciTech Connect (OSTI)

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

1994-03-01T23:59:59.000Z

253

Assessing compliance with the EPA high-level waste standard: an overview  

SciTech Connect (OSTI)

The US Environmental Protection Agency has set a standard for the performance of geologic repositories for the disposal of high-level radioactive waste. The standard is divided into several sections, including a section on containment requirements. The containment requirement is probabilistic, in that it allows certain small amounts of radioactive waste to be released at high probabilities and larger amounts to be released at lower probabilities. The US Nuclear Regulatory Commission is responsible for implementing the standard. Implementation of the standard will probably involve development and screening of scenarios, assignment of probabilities to the scenarios, determination of consequences of the scenarios, and analysis of uncertainties. Scenario development consists of first, identifying events and processes that could initiate waste releases or affect waste transport, and second, combining the events and processes in physically reasonable ways. Scenarios can be screened on the basis of low probabilities or consequences. Consequences of scenarios are estimated using a series of models that simulate the movement of radionuclides out of the waste package and underground facility and the transport of the radionuclides by ground water or other means to the accessible environment. Sensitivity and uncertainty analysis examines the sources and effects of uncertainties on the calculations. This document uses a simple example to illustrate techniques for the implementation of the standard.

Hunter, R.L.; Cranwell, R.M.; Chu, M.S.Y.

1986-10-01T23:59:59.000Z

254

Waste component recycle, treatment, and disposal integrated demonstration (WeDID) nuclear weapon dismantlement activities  

SciTech Connect (OSTI)

One of the drivers in the dismantlement and disposal of nuclear weapon components is Envirorunental Protection Agency (EPA) guidelines. The primary regulatory driver for these components is the Resource Conservation Recovery Act (RCRA). Nuclear weapon components are heterogeneous and contain a number of hazardous materials including heavy metals, PCB`S, selfcontained explosives, radioactive materials, gas-filled tubes, etc. The Waste Component Recycle, Treatment, Disposal and Integrated Demonstration (WeDID) is a Department of Energy (DOE) Environmental Restoration and Waste Management (ERWM) sponsored program. It also supports DOE Defense Program (DP) dismantlement activities. The goal of WeDID is to demonstrate the end-to-end disposal process for Sandia National Laboratories designed nuclear weapon components. One of the primary objectives of WeDID is to develop and demonstrate advanced system treatment technologies that will allow DOE to continue dismantlement and disposal unhindered even as environmental regulations become more stringent. WeDID is also demonstrating waste minimization techniques by recycling a significant weight percentage of the bulk/precious metals found in weapon components and by destroying the organic materials typically found in these components. WeDID is concentrating on demonstrating technologies that are regulatory compliant, are cost effective, technologically robust, and are near-term to ensure the support of DOE dismantlement time lines. The waste minimization technologies being demonstrated by WeDID are cross cutting and should be able to support a number of ERWM programs.

Wheelis, W.T.

1993-04-12T23:59:59.000Z

255

Waste component recycle, treatment, and disposal integrated demonstration (WeDID) nuclear weapon dismantlement activities  

SciTech Connect (OSTI)

One of the drivers in the dismantlement and disposal of nuclear weapon components is Envirorunental Protection Agency (EPA) guidelines. The primary regulatory driver for these components is the Resource Conservation Recovery Act (RCRA). Nuclear weapon components are heterogeneous and contain a number of hazardous materials including heavy metals, PCB'S, selfcontained explosives, radioactive materials, gas-filled tubes, etc. The Waste Component Recycle, Treatment, Disposal and Integrated Demonstration (WeDID) is a Department of Energy (DOE) Environmental Restoration and Waste Management (ERWM) sponsored program. It also supports DOE Defense Program (DP) dismantlement activities. The goal of WeDID is to demonstrate the end-to-end disposal process for Sandia National Laboratories designed nuclear weapon components. One of the primary objectives of WeDID is to develop and demonstrate advanced system treatment technologies that will allow DOE to continue dismantlement and disposal unhindered even as environmental regulations become more stringent. WeDID is also demonstrating waste minimization techniques by recycling a significant weight percentage of the bulk/precious metals found in weapon components and by destroying the organic materials typically found in these components. WeDID is concentrating on demonstrating technologies that are regulatory compliant, are cost effective, technologically robust, and are near-term to ensure the support of DOE dismantlement time lines. The waste minimization technologies being demonstrated by WeDID are cross cutting and should be able to support a number of ERWM programs.

Wheelis, W.T.

1993-04-12T23:59:59.000Z

256

Secondary wastes and high explosive residues generated during production of main high explosive charges for nuclear weapons  

SciTech Connect (OSTI)

This study identifies the sources of high-explosive (HE) residues and hazardous and nonhazardous wastes generated during the production of the main HE charges for nuclear weapons, and estimates their quantities and characteristics. The results can be used as a basis for design of future handling and treatment systems for solid and liquid HE residues and wastes at any proposed new HE production facilities. This paper outlines a general methodology for documenting and estimating the volumes and characteristics of the solid and liquid HE residues and hazardous and nonhazardous wastes. We prepared volume estimates by applying this method to actual past Pantex plant HE production operations. To facilitate the estimating, we separated the HE main-charge production process into ten discrete unit operations and four support operations, and identified the corresponding solid and liquid HE residues and waste quantities. Four different annual HE main-charge production rates of 100, 500, 1000, and 2000 HE units/yr were assumed to develop the volume estimates and to establish the sensitivity of the estimates to HE production rates. The total solids (HE residues and hazardous and nonhazardous wastes) estimated range from 800 to 2800 ft{sup 3}/yr and vary uniformly with the assumed HE production rate. The total liquids estimated range from 73,000 to 1,448,000 gal/yr and also vary uniformly with the assumed production rate.

Jardine, L.J.; McGee, J.T.

1994-02-01T23:59:59.000Z

257

Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)  

SciTech Connect (OSTI)

The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

Burgard, K.C.

1998-04-09T23:59:59.000Z

258

Geological Constraints on High-Level Nuclear Waste Disposal and their Relationship to Possible  

E-Print Network [OSTI]

to Possible Long Term Storage Solutions- A Case Study of the Yucca Mountain Project Teresa Dunn 2013 #12;Dunn systems and geologic composition in the selection and development of a secure, long-term storage facilityDunn 1 Geological Constraints on High-Level Nuclear Waste Disposal and their Relationship

Polly, David

259

HLW-OVP-97-0068 High Level Waste Management Division High-Level...  

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

not exceed the applicable concentration limits for Class C low-level waste as set out in 10 CFR Part 61; and (c) will be managed, pursuant to the Atomic Energy Act, so that safety...

260

Conversion of an aluminosilicate-based waste material to high-value efficient adsorbent  

Science Journals Connector (OSTI)

Abstract The recycling of waste printed circuit boards (PCBs) has become one of the global challenges in the technological era. The colossal volume of waste PCB generated annually coupled with its toxic nature and the existence of highly-precious metals in its composition intensifies the problems associated with waste PCB management and recycling. The two prevalent waste management options, landfill disposal and incineration, are being phased out for this special waste as a result of public health concerns. Hence, in the past few decades, several PCB recycling schemes are being introduced. The most efficient and environmentally-sound practice for waste PCB recycling has been the separation of metallic and nonmetallic fraction of \\{PCBs\\} by extensively-studied physico-mechanical approaches. Although the metallic fraction can be directly rendered into the market due to its high value, the nonmetallic fraction (NMF) is either disposed of in landfills causing secondary pollution or used as a low-value filler with the sole purpose of its safe disposal. This study presents a brief overview of the utilization of NMF as a filler in various industries. The main objective of the present review is to thoroughly examine the novel, highly efficient application of NMF as precursor for the production of a mesoporous structured adsorbent and its application in the removal of a myriad of heavy metals in single- and multi-component systems. In addition, the effects of the operational parameters on the adsorption behavior of the adsorbent material have been provided. Moreover, a comprehensive overview of the adsorption system modelling for single and binary-component systems for this novel material has been compiled.

Pejman Hadi; Chao Ning; Weiyi Ouyang; Carol Sze Ki Lin; Chi-Wai Hui; Gordon McKay

2014-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

Engineering development of a lightweight high-pressure scarifier for tank waste retrieval  

SciTech Connect (OSTI)

The Retrieval Process Development and Enhancements Program (RPD&E) is sponsored by the U.S. Department of Energy Tanks Focus Area to investigate existing and emerging retrieval processes suitable for the retrieval of high-level radioactive waste inside underground storage tanks. This program, represented by industry, national laboratories, and academia, seeks to provide a technical and cost basis to support site-remediation decisions. Part of this program has involved the development of a high-pressure waterjet dislodging system and pneumatic conveyance integrated as a scarifier. Industry has used high-pressure waterjet technology for many years to mine, cut, clean, and scarify materials with a broad range of properties. The scarifier was developed as an alternate means of retrieving waste inside Hanford single-shell tanks, particularly hard, stubborn waste. Testing of the scarifier has verified its ability to retrieve a wide range of tank waste ranging from extremely hard waste that is resistant to other dislodging means to soft sludge and even supernatant fluid. Since the scarifier expends water at a low rate and recovers most of the water as it is used, the scarifier is well suited for retrieval of tanks that leak and cannot be safely sluiced or applications where significant waste dilution is not acceptable. Although the original scarifier was effective, it became evident that a lighter, more compact version that would be compatible with light weight deployment systems under development, such as the Light Duty Utility Arm, was needed. At the end of FY 95, the Light Weight Scarifier (LWS) was designed to incorporate the features of the original scarifier in a smaller, lighter end effector. During FY 96, the detailed design of the LWS was completed and two prototypes were fabricated.

Hatchell, B.K.

1997-09-01T23:59:59.000Z

262

Waste Inspection Tomography (WIT)  

SciTech Connect (OSTI)

Waste Inspection Tomography (WIT) provides mobile semi-trailer mounted nondestructive examination (NDE) and assay (NDA) for nuclear waste drum characterization. WIT uses various computed tomography (CT) methods for both NDE and NDA of nuclear waste drums. Low level waste (LLW), transuranic (TRU), and mixed radioactive waste can be inspected and characterized without opening the drums. With externally transmitted x-ray NDE techniques, WIT has the ability to identify high density waste materials like heavy metals, define drum contents in two- and three-dimensional space, quantify free liquid volumes through density and x-ray attenuation coefficient discrimination, and measure drum wall thickness. With waste emitting gamma-ray NDA techniques, WIT can locate gamma emitting radioactive sources in two- and three-dimensional space, identify gamma emitting isotopic species, identify the external activity levels of emitting gamma-ray sources, correct for waste matrix attenuation, provide internal activity approximations, and provide the data needed for waste classification as LLW or TRU. The mobile feature of WIT allows inspection technologies to be brought to the nuclear waste drum storage site without the need to relocate drums for safe, rapid, and cost-effective characterization of regulated nuclear waste. The combination of these WIT characterization modalities provides the inspector with an unprecedented ability to non-invasively characterize the regulated contents of waste drums as large as 110 gallons, weighing up to 1,600 pounds. Any objects that fit within these size and weight restrictions can also be inspected on WIT, such as smaller waste bags and drums that are five and thirty-five gallons.

Bernardi, R.T.

1995-12-01T23:59:59.000Z

263

Annual summary of Immobilized Low-Activity Waste (ILAW) Performance Assessment for 2003 Incorporating the Integrated Disposal Facility Concept  

SciTech Connect (OSTI)

To Erik Olds 09/30/03 - An annual summary of the adequacy of the Hanford Immobilized Low-Activity Tank Waste Performance Assessment (ILAW PA) is necessary in each year in which a full performance assessment is not issued.

MANN, F M

2003-09-01T23:59:59.000Z

264

Potential Application Of Radionuclide Scaling Factors To High Level Waste Characterization  

SciTech Connect (OSTI)

Production sources, radiological properties, relative solubilities in waste, and laboratory analysis techniques for the forty-five radionuclides identified in Hanford?s Waste Treatment and Immobilization Plant (WTP) Feed Acceptance Data Quality Objectives (DQO) document are addressed in this report. Based on Savannah River Site (SRS) experience and waste characteristics, thirteen of the radionuclides are judged to be candidates for potential scaling in High Level Waste (HLW) based on the concentrations of other radionuclides as determined through laboratory measurements. The thirteen radionuclides conducive to potential scaling are: Ni-59, Zr-93, Nb-93m, Cd-113m, Sn-121m, Sn-126, Cs-135, Sm-151, Ra-226, Ra-228, Ac-227, Pa-231, and Th-229. The ability to scale radionuclides is useful from two primary perspectives: 1) it provides a means of checking the radionuclide concentrations that have been determined by laboratory analysis; and 2) it provides a means of estimating radionuclide concentrations in the absence of a laboratory analysis technique or when a complex laboratory analysis technique fails. Along with the rationale for identifying and applying the potential scaling factors, this report also provides examples of using the scaling factors to estimate concentrations of radionuclides in current SRS waste and into the future. Also included in the report are examples of independent laboratory analysis techniques that can be used to check results of key radionuclide analyses. Effective utilization of radionuclide scaling factors requires understanding of the applicable production sources and the chemistry of the waste. As such, the potential scaling approaches identified in this report should be assessed from the perspective of the Hanford waste before reaching a decision regarding WTP applicability.

Reboul, S. H.

2013-09-30T23:59:59.000Z

265

High Level Overview of DOE Biomass Logistics II Project Activities...  

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

High Level Overview of DOE Biomass Logistics II Project Activities High Level Overview of DOE Biomass Logistics II Project Activities Breakout Session 1B-Integration of Supply...

266

Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses  

SciTech Connect (OSTI)

Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30 deg. to 60 deg. C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under {gamma}-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. (authors)

Pankov, Alexey S.; Ojovan, Michael I. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield, S1 3JD (United Kingdom); Batyukhnova, Olga G. [International Education Training Centre, SUE SIA 'Radon', The 7-th Rostovsky Lane 2/14, Moscow, 119121 (Russian Federation); Lee, William E. [Department of Materials, Imperial College London, South Kensington Campus, Exhibition Road, London, SW7 2AZ (United Kingdom)

2007-07-01T23:59:59.000Z

267

Assessment of commercially available ion exchange materials for cesium removal from highly alkaline wastes  

SciTech Connect (OSTI)

Approximately 61 million gallons of nuclear waste generated in plutonium production, radionuclide removal campaigns, and research and development activities is stored on the Department of Energy`s Hanford Site, near Richland, Washington. Although the pretreatment process and disposal requirements are still being defined, most pretreatment scenarios include removal of cesium from the aqueous streams. In many cases, after cesium is removed, the dissolved salt cakes and supernates can be disposed of as LLW. Ion exchange has been a leading candidate for this separation. Ion exchange systems have the advantage of simplicity of equipment and operation and provide many theoretical stages in a small space. The organic ion exchange material Duolite{trademark} CS-100 has been selected as the baseline exchanger for conceptual design of the Initial Pretreatment Module (IPM). Use of CS-100 was chosen because it is considered a conservative, technologically feasible approach. During FY 96, final resin down-selection will occur for IPM Title 1 design. Alternate ion exchange materials for cesium exchange will be considered at that time. The purpose of this report is to conduct a search for commercially available ion exchange materials which could potentially replace CS-100. This report will provide where possible a comparison of these resin in their ability to remove low concentrations of cesium from highly alkaline solutions. Materials which show promise can be studied further, while less encouraging resins can be eliminated from consideration.

Brooks, K.P.; Kim, A.Y.; Kurath, D.E.

1996-04-01T23:59:59.000Z

268

High-level waste canister storage final design, installation, and testing. Topical report  

SciTech Connect (OSTI)

This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

1998-04-01T23:59:59.000Z

269

Ceramic stabilization of hazardous wastes: a high performance room temperature process  

SciTech Connect (OSTI)

ANL has developed a room-temperature process for converting hazardous materials to a ceramic structure. It is similar to vitrification but is achieved at low cost, similar to conventional cement stabilization. The waste constituents are both chemically stabilized and physically encapsulated, producing very low leaching levels and the potential for delisting. The process, which is pH-insensitive, is ideal for inorganic sludges and liquids, as well as mixed chemical-radioactive wastes, but can also handle significant percentages of salts and even halogenated organics. High waste loadings are possible and densification occurs,so that volumes are only slightly increased and in some cases (eg, incinerator ash) are reduced. The ceramic product has strength and weathering properties far superior to cement products.

Maloney, M.D.

1996-10-01T23:59:59.000Z

270

Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage  

SciTech Connect (OSTI)

It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanisms have been identified that could, under certain conditions, cause failure of the SF cladding and/or failure of TRUW storage containers. Stress rupture limits for Zircaloy-clad SF in MRS range from 300 to 440/sup 0/C, based on limited data. Additional tests on irradiated Zircaloy (3- to 5-year duration) are needed to narrow this uncertainty. Cladding defect sizes could increase in air as a result of fuel density decreases due to oxidation. Oxidation tests (3- to 5-year duration) on SF are also needed to verify oxidation rates in air and to determine temperatures below which monitoring of an inert cover gas would not be required. Few, if any, changes in the physical state of HLW glass or canisters or their performance would occur under projected MRS conditions. The major uncertainty for HLW is in the heat transfer through cracked glass and glass devitrification above 500/sup 0/C. Additional study of TRUW is required. Some fraction of present TRUW containers would probably fail within the first 100 years of MRS, and some TRUW would be highly degraded upon retrieval, even in unfailed containers. One possible solution is the design of a 100-year container. 93 references, 28 figures, 17 tables.

Guenther, R.J.; Gilbert, E.R.; Slate, S.C.; Partain, W.L.; Divine, J.R.; Kreid, D.K.

1984-01-01T23:59:59.000Z

271

High-Level Waste Tank Cleaning and Field Characterization at the West Valley Demonstration Project  

SciTech Connect (OSTI)

The West Valley Demonstration Project (WVDP) is nearing completion of radioactive high-level waste (HLW) retrieval from its storage tanks and subsequent vitrification of the HLW into borosilicate glass. Currently, 99.5% of the sludge radioactivity has been recovered from the storage tanks and vitrified. Waste recovery of cesium-137 (Cs-137) adsorbed on a zeolite media during waste pretreatment has resulted in 97% of this radioactivity being vitrified. Approximately 84% of the original 1.1 x 1018 becquerels (30 million curies) of radioactivity was efficiently vitrified from July 1996 to June 1998 during Phase I processing. The recovery of the last 16% of the waste has been challenging due to a number of factors, primarily the complex internal structural support system within the main 2.8 million liter (750,000 gallon) HLW tank designated 8D-2. Recovery of this last waste has become exponentially more challenging as less and less HLW is available to mobilize and transfer to the Vitrification Facility. This paper describes the progressively more complex techniques being utilized to remove the final small percentage of radioactivity from the HLW tanks, and the multiple characterization technologies deployed to determine the quantity of Cs-137, strontium-90 (Sr-90), and alpha-transuranic (alpha-TRU) radioactivity remaining in the tanks.

Drake, J. L.; McMahon, C. L.; Meess, D. C.

2002-02-26T23:59:59.000Z

272

Cold Crucible Induction Melting Technology for Vitrification of High Level Waste: Development and Status in India  

SciTech Connect (OSTI)

Cold crucible induction melting is globally emerging as an alternative technology for the vitrification of high level radioactive waste. The new technology offers several advantages such as high temperature availability with long melter life, high waste loading, high specific capacity etc. Based on the laboratory and bench scale studies, an engineering scale cold crucible induction melter was locally developed in India. The melter was operated continuously to assess its performance. The electrical and thermal efficiencies were found to be in the range of 70-80 % and 10-20 % respectively. Glass melting capacities up to 200 kg m{sup -2} hr{sup -1} were accomplished using the ESCCIM. Industrially adaptable melter operating procedures for start-up, melting and pouring operations were established (author)

Sugilal, G.; Sengar, P.B.S. [Nuclear Recycle Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

2008-07-01T23:59:59.000Z

273

Treatment of Liquid Radioactive Waste with High Salt Content by Colloidal Adsorbents - 13274  

SciTech Connect (OSTI)

Treatment processes have been fully developed for most of the liquid radioactive wastes generated during the operation of nuclear power plants. However, a process for radioactive liquid waste with high salt content, such as waste seawater generated from the unexpected accident at nuclear power station, has not been studied extensively. In this study, the adsorption efficiencies of cesium (Cs) and strontium (Sr) in radioactive liquid waste with high salt content were investigated using several types of zeolite with different particle sizes. Synthesized and commercial zeolites were used for the treatment of simulated seawater containing Cs and Sr, and the reaction kinetics and adsorption capacities of colloidal zeolites were compared with those of bulk zeolites. The experimental results demonstrated that the colloidal adsorbents showed fast adsorption kinetic and high binding capacity for Cs and Sr. Also, the colloidal zeolites could be successfully applied to the static adsorption condition, therefore, an economical benefit might be expected in an actual processes where stirring is not achievable. (authors)

Lee, Keun-Young; Chung, Dong-Yong; Kim, Kwang-Wook; Lee, Eil-Hee; Moon, Jei-Kwon [Korea Atomic Energy Research Institute - KAERI, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)] [Korea Atomic Energy Research Institute - KAERI, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2013-07-01T23:59:59.000Z

274

River Protection Project (RPP) Immobilized Low Activity Waste (ILAW) Disposal Plan  

SciTech Connect (OSTI)

This document replaces HNF-1517, Rev 2 which is deleted. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract change and associated DOE/ORP guidance. In addition it incorporates the cancellation of Project W-465, Grout Facility, and the associated modifications to Project W-520, Immobilized High-Level Waste Disposal Facility. It also includes document format changes and section number modifications consistent with CH2M HILL Hanford Group, Inc. procedures.

BRIGGS, M.G.

2000-09-22T23:59:59.000Z

275

DOE/EIS-0287 Idaho High-Level Waste & Facilities Disposition Draft Environmental Impact Statement (December 1999)  

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

HLW & FD EIS HLW & FD EIS 3-13 DOE/EIS-0287D Calcine storag e i n b i n s ets Calcine storag e i n b i n s et s Cesium ion exchange & grouting Cesium ion exchange & grouting NWCF* NWCF* Calcine Mixed transuranic waste/SBW Mixed transuranic waste/NGLW Low-level waste disposa l*** disposa l*** Tank heels Transuranic waste (from tank heels) * * * * Mixed transuranic waste/ NGLW Mixed transuranic waste/ NGLW M i x e d t r a nsuran ic w a s t e / M i x e d t r a nsuran ic w a s t e / S B W s t o rage in Ta n k F a r m S B W s t o rage in Ta n k F a r m Low-leve l waste Low-leve l waste FIGURE 3-2. Continued Current Operations Alternative. LEGEND * Including high-temperature and maximum achievable control technology upgrades. Mixed transuranic waste/ newly generated liquid waste New Waste Calcining Facility ** Calcine would be transferred from bin set #1 to bin set #6 or #7.

276

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers  

SciTech Connect (OSTI)

Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel {approx} Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs.

Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-08-01T23:59:59.000Z

277

Secondary wastes and high explosive residues generated during production of main high explosive charges for nuclear weapons. Revision 1  

SciTech Connect (OSTI)

This study identifies the sources of high-explosive (HE) residues and hazardous and nonhazardous wastes generated during the production of the main HE charges for nuclear weapons, and estimates their quantities and characteristics. The results can be used as a basis for design of future handling and treatment systems for solid and liquid HE residues and wastes at any proposed new HE production facilities. This paper outlines a general methodology for documenting and estimating the volumes and characteristics of the solid and liquid HE residues and hazardous and nonhazardous wastes. To facilitate the estimating, we separated the HE main-charge production process into ten discrete unit operations and four support operations, and identified the corresponding solid and liquid HE residues and waste quantities. Four different annual HE main-charge production rates of 100, 500, 1000, and 2000 HE units/yr were assumed to develop the volume estimates and to establish the sensitivity of the estimates to HE production rates. The total solids (HE residues and hazardous and nonhazardous wastes) estimated range from 800 to 2800 ft{sup 3}/yr and vary uniformly with the assumed HE production rate. The total liquids estimated range from 73,000 to 1,448.000 gal/yr and also vary uniformly with the assumed production rate. Of the estimated solids, the hazardous wastes (e.g., electrical vehicle batteries and light tubes) were about 2% of the total volumes. The generation of solid HE residues varied uniformly with the HE production rates and ranged from about 20% of the total solids volume for the 100 HE units/yr case to about 60% for the 2000 units/yr case. The HE machining operations generated 60 to 80% of the total solid HE residues, depending on the assumed production rate, and were also the sources of the most concentrated HE residues.

Jardine, L.J.; McGee, J.T.

1995-01-01T23:59:59.000Z

278

Waste degradation and mobilization in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste  

Science Journals Connector (OSTI)

Abstract This paper summarizes modeling of waste degradation and mobilization in performance assessments (PAs) conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. As understanding of the Yucca Mountain disposal system increased, the waste degradation module, or succinctly called the source-term, evolved from initial assumptions in 1984 to results based on process modeling in 2008. In early PAs, waste degradation had significant influence on calculated behavior but as the robustness of the waste container was increased and modeling of the container degradation improved, waste degradation had much less influence in later PAs. The variation of dissolved concentrations of radionuclides progressed from simple probability distributions in early \\{PAs\\} to functions dependent upon water chemistry in later PAs. Also, transport modeling of radionuclides in the waste, container, and invert were added in 1995; and, colloid-facilitated transport of radionuclides was added in 1998.

Rob P. Rechard; Christine T. Stockman

2014-01-01T23:59:59.000Z

279

Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333  

SciTech Connect (OSTI)

The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion adsorption chemicals, solid-liquid separation methods, and achievable decontamination factors. Results of the radionuclide removal testing indicate that the radionuclides, including Tc-99, can be removed with inorganic sorbents and precipitating agents. Evaporation test results indicate that the simulant can be evaporated to fairly high concentration prior to formation of appreciable solids, but corrosion has not yet been examined.

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.; Taylor-Pashow, Kathryn M.; Adamson, Duane J.; Crawford, Charles L.; Morse, Megan M.

2014-01-07T23:59:59.000Z

280

Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics. Revision 1  

SciTech Connect (OSTI)

The Department of Energy`s (DOE`s) planning for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of the waste. This report estimates volumes, radionuclide activities, and waste forms of GTCC LLW to the year 2035. It groups the waste into four categories, representative of the type of generator or holder of the waste: Nuclear Utilities, Sealed Sources, DOE-Held, and Other Generator. GTCC LLW includes activated metals (activation hardware from reactor operation and decommissioning), process wastes (i.e., resins, filters, etc.), sealed sources, and other wastes routinely generated by users of radioactive material. Estimates reflect the possible effect that packaging and concentration averaging may have on the total volume of GTCC LLW. Possible GTCC mixed LLW is also addressed. Nuclear utilities will probably generate the largest future volume of GTCC LLW with 65--83% of the total volume. The other generators will generate 17--23% of the waste volume, while GTCC sealed sources are expected to contribute 1--12%. A legal review of DOE`s obligations indicates that the current DOE-Held wastes described in this report will not require management as GTCC LLW because of the contractual circumstances under which they were accepted for storage. This report concludes that the volume of GTCC LLW should not pose a significant management problem from a scientific or technical standpoint. The projected volume is small enough to indicate that a dedicated GTCC LLW disposal facility may not be justified. Instead, co-disposal with other waste types is being considered as an option.

Not Available

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

National Low-Level Waste Management Program final summary report of key activities and accomplishments for fiscal year 1997  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has responsibilities under the Low-Level Radioactive Waste Policy Amendments Act of 1985 to assist states and compacts in their siting and licensing efforts for low-level radioactive waste disposal facilities. The National Low-Level Waste Management Program (NLLWMP) is the element of the DOE that performs the key support activities under the Act. The NLLWMP`s activities are driven by the needs of the states and compacts as they prepare to manage their low-level waste under the Act. Other work is added during the fiscal year as necessary to accommodate new requests brought on by status changes in states` and compacts` siting and licensing efforts. This report summarizes the activities and accomplishments of the NLLWMP during FY 1997.

Rittenberg, R.B.

1998-03-01T23:59:59.000Z

282

High Level Waste System Impacts from Small Column Ion Exchange Implementation  

SciTech Connect (OSTI)

The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were combined into three batches for a total of about 3.2 million gallons of liquid waste. The chemical and radiological composition of these batches was estimated from the SpaceMan Plus{trademark} model using the same data set and assumptions as the baseline plans.

McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

2005-08-18T23:59:59.000Z

283

Activity of Microorganisms in Organic Waste Disposal: IV. Bio-Calculations  

Science Journals Connector (OSTI)

...characteristics of the sludge system as previously...example, an activated sludge from a food processing...its high activity and gas production. As a result, the maximum attainable sludge concentration was 0...solution in microscopic bubbles. These fine bubbles...

W. Wesley Eckenfelder Jr.; Nandor Porges

1957-05-01T23:59:59.000Z

284

Waste Heat Recovery from High Temperature Off-Gases from Electric Arc Furnace  

SciTech Connect (OSTI)

This article presents a study and review of available waste heat in high temperature Electric Arc Furnace (EAF) off gases and heat recovery techniques/methods from these gases. It gives details of the quality and quantity of the sensible and chemical waste heat in typical EAF off gases, energy savings potential by recovering part of this heat, a comprehensive review of currently used waste heat recovery methods and potential for use of advanced designs to achieve a much higher level of heat recovery including scrap preheating, steam production and electric power generation. Based on our preliminary analysis, currently, for all electric arc furnaces used in the US steel industry, the energy savings potential is equivalent to approximately 31 trillion Btu per year or 32.7 peta Joules per year (approximately $182 million US dollars/year). This article describes the EAF off-gas enthalpy model developed at Oak Ridge National Laboratory (ORNL) to calculate available and recoverable heat energy for a given stream of exhaust gases coming out of one or multiple EAF furnaces. This Excel based model calculates sensible and chemical enthalpy of the EAF off-gases during tap to tap time accounting for variation in quantity and quality of off gases. The model can be used to estimate energy saved through scrap preheating and other possible uses such as steam generation and electric power generation using off gas waste heat. This article includes a review of the historical development of existing waste heat recovery methods, their operations, and advantages/limitations of these methods. This paper also describes a program to develop and test advanced concepts for scrap preheating, steam production and electricity generation through use of waste heat recovery from the chemical and sensible heat contained in the EAF off gases with addition of minimum amount of dilution or cooling air upstream of pollution control equipment such as bag houses.

Nimbalkar, Sachin U [ORNL; Thekdi, Arvind [E3M Inc; Keiser, James R [ORNL; Storey, John Morse [ORNL

2014-01-01T23:59:59.000Z

285

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7  

SciTech Connect (OSTI)

This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms.

Not Available

1994-04-01T23:59:59.000Z

286

HANFORD MEDIUM & LOW CURIE WASTE PRETREATMENT PROJECT PHASE 1 LAB REPORT  

SciTech Connect (OSTI)

A fractional crystallization (FC) process is being developed to supplement tank waste pretreatment capabilities provided by the Waste Treatment and Immobilization Plant (WTP). FC can process many tank wastes, separating wastes into a low-activity fraction (LAW) and high-activity fraction (HLW). The low-activity fraction can be immobilized in a glass waste form by processing in the bulk vitrification (BV) system.

HAMILTON, D.W.

2006-01-30T23:59:59.000Z

287

Chemical and biological systems for regenerating activated carbon contaminated with high explosives  

SciTech Connect (OSTI)

Activated carbon has been used as a substrate for efficiently removing high explosives (HEs) from aqueous and gaseous waste streams. Carbon that is saturated with HEs, however, constitutes a solid waste and is currently being stored because appropriate technologies for its treatment are not available. Because conventional treatment strategies (i.e., incineration, open burning) are not safe or will not be in compliance with future regulations, new and cost-effective methods are required for the elimination of this solid waste. Furthermore, because the purchase of activated carbon and its disposal after loading with HEs will be expensive, an ideal treatment method would result in the regeneration of the carbon thereby permitting its reuse. Coupling chemical and biological treatment systems, such as those described below, will effectively meet these technical requirements. The successful completion of this project will result in the creation of engineered commercial systems that will present safe and efficient methods for reducing the quantities of HE-laden activated carbon wastes that are currently in storage or are generated as a result of demilitarization activities. Biological treatment of hazardous wastes is desirable because the biodegradation process ultimately leads to the mineralization (e.g., conversion to carbon dioxide, nitrogen gas, and water) of parent compounds and has favorable public acceptance. These methods will also be cost- effective because they will not require large expenditures of energy and will permit the reuse of the activated carbon. Accordingly, this technology will have broad applications in the private sector and will be a prime candidate for technology transfer.

Knezovich, J.P.; Daniels, J.I. [Lawrence Livermore National Lab., CA (United States); Stenstrom, M.K.; Heilmann, H.M. [Univ. of California, Los Angeles, CA (United States). Civil and Engineering Dept.

1994-12-01T23:59:59.000Z

288

High energy activation data library (HEAD-2009)  

SciTech Connect (OSTI)

A proton activation data library for 682 nuclides from 1 H to 210Po in the energy range from 150 MeV up to 1 GeV was developed. To calculate proton activation data, the MCNPX 2.6.0 and CASCADE/INPE codes were chosen. Different intranuclear cascade, preequilibrium, and equilibrium nuclear reaction models and their combinations were used. The optimum calculation models have been chosen on the basis of statistical correlations for calculated and experimental proton data taken from the EXFOR library of experimental nuclear data. All the data are written in ENDF-6 format. The library is called HEPAD-2008 (High-Energy Proton Activation Data). A revision of IEAF-2005 neutron activation data library has been performed. A set of nuclides for which the cross-section data can be (and were) updated using more modern and improved models is specified, and the corresponding calculations have been made in the present work. The new version of the library is called IEAF-2009. The HEPAD-2008 and IEAF-2009 are merged to the final HEAD-2009 library.

Mashnik, Stepan G [Los Alamos National Laboratory; Korovin, Yury A [NON LANL; Natalenko, Anatoly A [NON LANL; Konobeyev, Alexander Yu [NON LANL; Stankovskiy, A Yu [NON LANL

2010-01-01T23:59:59.000Z

289

TWRS Retrieval and Storage Mission and Immobilized Low Activity Waste (ILAW) Disposal Plan  

SciTech Connect (OSTI)

This project plan has a twofold purpose. First, it provides a waste stream project plan specific to the River Protection Project (RPP) (formerly the Tank Waste Remediation System [TWRS] Project) Immobilized Low-Activity Waste (LAW) Disposal Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-90-01 (Ecology et al. 1994) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan (Ecology et al. 1998). Second, it provides an upper tier document that can be used as the basis for future subproject line-item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 [DOE 1992] and 430.1 [DOE 1995a]). The format and content of this project plan are designed to accommodate the requirements mentioned by the Tri-Party Agreement and the DOE orders. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

BURBANK, D.A.

1999-09-01T23:59:59.000Z

290

Round-robin testing of a reference glass for low-activity waste forms  

SciTech Connect (OSTI)

A round robin test program was conducted with a glass that was developed for use as a standard test material for acceptance testing of low-activity waste glasses made with Hanford tank wastes. The glass is referred to as the low-activity test reference material (LRM). The program was conducted to measure the interlaboratory reproducibility of composition analysis and durability test results. Participants were allowed to select the methods used to analyze the glass composition. The durability tests closely followed the Product Consistency Test (PCT) Method A, except that tests were conducted at both 40 and 90 C and that parallel tests with a reference glass were not required. Samples of LRM glass that had been crushed, sieved, and washed to remove fines were provided to participants for tests and analyses. The reproducibility of both the composition and PCT results compare favorably with the results of interlaboratory studies conducted with other glasses. From the perspective of reproducibility of analysis results, this glass is acceptable for use as a composition standard for nonradioactive components of low-activity waste forms present at >0.1 elemental mass % and as a test standard for PCTS at 40 and 90 C. For PCT with LRM glass, the expected test results at the 95% confidence level are as follows: (1) at 40 C: pH = 9.86 {+-} 0.96; [B] = 2.30 {+-} 1.25 mg/L; [Na] = 19.7 {+-} 7.3 mg/L; [Si] = 13.7 {+-} 4.2 mg/L; and (2) at 90 C: pH = 10.92 {+-} 0.43; [B] = 26.7 {+-} 7.2 mg/L; [Na] = 160 {+-} 13 mg/L; [Si] = 82.0 {+-} 12.7 mg/L. These ranges can be used to evaluate the accuracy of PCTS conducted at other laboratories.

Ebert, W. L.; Wolf, S. F.

1999-12-06T23:59:59.000Z

291

World first in high level waste vitrification - A review of French vitrification industrial achievements  

SciTech Connect (OSTI)

AREVA has more than 30 years experience in operating industrial HLW (High Level radioactive Waste) vitrification facilities (AVM - Marcoule Vitrification Facility, R7 and T7 facilities). This vitrification technology was based on borosilicate glasses and induction-heating. AVM was the world's first industrial HLW vitrification facility to operate in-line with a reprocessing plant. The glass formulation was adapted to commercial Light Water Reactor fission products solutions, including alkaline liquid waste concentrates as well as platinoid-rich clarification fines. The R7 and T7 facilities were designed on the basis of the industrial experience acquired in the AVM facility. The AVM vitrification process was implemented at a larger scale in order to operate the R7 and T7 facilities in-line with the UP2 and UP3 reprocessing plants. After more than 30 years of operation, outstanding record of operation has been established by the R7 and T7 facilities. The industrial startup of the CCIM (Cold Crucible Induction Melter) technology with enhanced glass formulation was possible thanks to the close cooperation between CEA and AREVA. CCIM is a water-cooled induction melter in which the glass frit and the waste are melted by direct high frequency induction. This technology allows the handling of highly corrosive solutions and high operating temperatures which permits new glass compositions and a higher glass production capacity. The CCIM technology has been implemented successfully at La Hague plant.

Brueziere, J.; Chauvin, E. [AREVA, 1 place Jean Millier, 92084 Paris La Defense (France); Piroux, J.C. [Joint Vitrification Laboratory - LCV, Marcoule, BP171, 30207 Bagnols sur Ceze (France)

2013-07-01T23:59:59.000Z

292

E-Print Network 3.0 - activity waste ilaw Sample Search Results  

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

Satellite Accumulation Areas (SAAs) All Hazardous waste generated... and California state regulations. All waste that is ... Source: Lawrence Berkeley National Laboratory, Ion Beam...

294

Options for Determining Equivalent MHTM (Metric Tons of Heavy Metal) for DOE High Level Waste  

SciTech Connect (OSTI)

Section 114(d) of the Nuclear Waste Policy Act of 1982, as amended (NWPA), limits the overall capacity of the first repository to 70,000 metric tons of heavy metal (MTHM). Current DOE policy is to allocate DOE spent fuel and high-level waste (HLW) at 10 percent of the total, or 7,000 MTHM. For planning purposes, 4,667 MTHM will be allocated for HLW. While the NWPA provides a technical basis for determining the MTHM equivalence of HLW, it does not address the significant technical differences between DOE HLW and commercial spent nuclear fuel (SNF). Although more than 170,000 MTHM of DOE fuel has been reprocessed to produce the inventory of HLW, the amount of radioactive waste generated per metric ton of DOE fuel is only a few percent of that in a metric ton of commercial fuel. This study compares the results of four methods for determining the MTHM equivalence of DOE HLW. These methods include (1) using the actual weight of heavy metal in reprocessed DOE fuel, (2) assuming the historical equivalence of 0.5 MTHM/canister of vitrified DOE HLW, (3) comparing the total radioactivity in DOE HLW to the radioactivity of commercial SNF, and (4) comparing the total radiotoxicity of DOE HLW, as defined for those radionuclides identified in 10 CFR 20, with SNF at 1,000 and 10,000 years. This study concludes that either of the last two options would meet Congress’s stated purposes of the NWPA, which are to (1) provide "reasonable assurance that the public and the environment will be adequately protected from the hazards posed by high-level radioactive waste and such spent nuclear fuel as may be disposed of in a repository", and (2) to "define Federal policy for the disposal of such waste and spent fuel".

Knecht, Dieter August; Valentine, James Henry; Luptak, Alan Jay; Staiger, Merle Daniel; Loo, Henry Hung Yiu; Wichmann, Thomas Leonard

1999-04-01T23:59:59.000Z

295

Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105 And AN-103) By Fluidized Bed Steam Reformation  

SciTech Connect (OSTI)

One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

Jantzen, Carol; Herman, Connie; Crawford, Charles; Bannochie, Christopher; Burket, Paul; Daniel, Gene; Cozzi, Alex; Nash, Charles; Miller, Donald; Missimer, David

2014-01-10T23:59:59.000Z

296

Dedicated-site, interim storage of high-level nuclear waste as part of the management system  

Science Journals Connector (OSTI)

...develop permanent repositories. A systems approach...reprocessed nuclear wastes from fuel rods...In- teragency Review Group (1) and...high-level wastes, including both...mined geologic repositories for permanent...Under current plans for mined geologic...and performance standards of permanent...

E-an Zen

1980-01-01T23:59:59.000Z

297

Title: An Advanced Solution for the Storage, Transportation and Disposal of Vitrified High Level Waste  

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

Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 1 AN ADVANCED SOLUTION FOR THE STORAGE, TRANSPORTATION AND DISPOSAL OF SPENT FUEL AND VITRIFIED HIGH LEVEL WASTE William J. Quapp Teton Technologies, Inc. 860 W. Riverview Dr. Idaho Falls, ID 83401 208-535-9001 ABSTRACT For future nuclear power deployment in the US, certain changes in the back end of the fuel cycle, i.e., disposal of high level waste and spent fuel, must become a real options. However, there exists another problem from the front end of the fuel cycle which has until recently, received less attention. Depleted uranium hexafluoride is a by-product of the enrichment process and has accumulated for over 50 years. It now represents a potential environmental problem. This paper describes a

298

Development of High-efficiency Thermoelectric Materials for Vehicle Waste Heat Utililization  

SciTech Connect (OSTI)

The goals of this . CRADA are: 1) Investigation of atomistic structure and nucleation of nanoprecipitates in (PbTe){sub I-x}(AgSbTe2){sub x} (LAST) system; and 2) Development of non-equilibrium synthesis of thermoelectric materials for waste heat recovery. We have made significant accomplishment in both areas. We studied the structure of LAST materials using high resolution imaging, nanoelectron diffraction, energy dispersive spectrum, arid electron energy loss spectrum, and observed a range of nanoparticles The results, published in J. of Applied Physics, provide quantitative structure information about nanoparticles, that is essential for the understanding of the origin of the high thermoelectric performance in this class of materials. We coordinated non-equilibrium synthesis and characterization of thermoelectric materials for waste heat recovery application. Our results, published in J. of Electronic Materials, show enhanced thermoelectric figure of merit and robust mechanical properties in bulk . filled skutterudites.

Li, Qiang

2009-04-30T23:59:59.000Z

299

Preliminary Study of Radioactive Waste Package Made of High-Strength and Ultra Low-Permeability Concrete for Geological Disposal of TRU Wastes  

SciTech Connect (OSTI)

We have been developing a radioactive waste package made of high-strength and ultra low-permeability concrete (HSULPC) for geological disposal of TRU wastes, which is expected to be much more impervious to water than conventional concrete. In this study, basic data for the HSULPC regarding its the impervious character and the thermodynamics during cement hydration were obtained through water permeability measurements using cold isostatic pressing (CIP) and adiabatic concrete hydration experiments, respectively. Then, a prediction tool to find concrete package construction conditions to avoid thermal cracking was developed, which could deal with coupled calculations of cement hydration, heat transfer, stress, and cracking. The developed tool was applied to HSULPC hydration on a small-scale cylindrical model to examine whether there was any effect on cracking which depended on the ratio of concrete cylinder thickness to its inner diameter. The results were compared to experiments. For concrete with a compressive strength of 200MPa, the water permeability coefficient was 4 x 10{sup 19} m/s. Dependences of activation energy and frequency factor on degree of cement hydration had a sharp peaking due to the nucleation rate-determining step, and a gradual increase region due to the diffusion rate-determining step. From analyses of the small-scale cylindrical model, dependences of the maximum principal stress on the radius were obtained. When the ratio of the concrete thickness to the heater diameter was around 1, the risk of cracking was predicted to be minimized. These numerical predictions from the developed tool were verified by experiments.

Matsuo, T.; Kawasaki, T.; Sakamoto, H.; Asano, E.; Takei, A.; Shibuya, K.; Katagiri, M.

2003-02-27T23:59:59.000Z

300

THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED BY MO K-EDGE X-RAY ABSORPTION  

E-Print Network [OSTI]

THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED of molybdenum in model UK high level nuclear waste glasses was investigated by X-ray Absorption Spectroscopy (XAS). Molybdenum K-edge XAS data were acquired from several inactive simulant high level nuclear waste

Sheffield, University of

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they are not comprehensive nor are they the most current set.
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301

Materials performance in a high-level radioactive waste vitrification system  

SciTech Connect (OSTI)

The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated.

Imrich, K.J.; Chandler, G.T.

1996-06-17T23:59:59.000Z

302

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

SciTech Connect (OSTI)

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

303

Application of Quantitative NDE Techniques to High Level Waste Storage Tanks  

SciTech Connect (OSTI)

As various issues make the continued usage of high-level waste storage tanks attractive, there is an increasing need to sharpen the assessment of their structural integrity. One aspect of a structural integrity program, nondestructive evaluation, is the focus of this paper. In September 2000, a program to support the sites was initiated jointly by Tanks Focus Area and Characterization, Monitoring, and Sensor Technologies Crosscutting Program of the Office of Environmental Management, Department of Energy (DOE). The vehicle was the Center for Nondestructive Evaluation, one of the National Science Foundation's Industry/University Cooperative Research Centers that is operated in close collaboration with the Ames Laboratory, USDOE. The support activities that have been provided by the center will be reviewed. Included are the organization of a series of annual workshops to allow the sites to share experiences and develop coordinated approaches to common problems, the development of an electronic source of relevant information, and assistance of the sites on particular technical problems. Directions and early results on some of these technical assistance projects are emphasized. Included are the discussion of theoretical analysis of ultrasonic wave propagation in curved plates to support the interpretation of tandem synthetic aperture focusing data to detect flaws in the knuckle region of double shell tanks; the evaluation of guided ultrasonic waves, excited by couplant free, electromagnetic acoustic transducers, to rapidly screen for inner wall corrosion in tanks; the use of spread spectrum techniques to gain information about the structural integrity of concrete domes; and the use of magnetic techniques to identify the alloys used in the construction of tanks.

Thompson, R. B.; Rehbein, D. K.; Bastiaans, G.; Terry, M.; Alers, R.

2002-02-25T23:59:59.000Z

304

DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS  

SciTech Connect (OSTI)

Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.

Jantzen, C

2006-01-06T23:59:59.000Z

305

22 - Radioactive waste disposal  

Science Journals Connector (OSTI)

Publisher Summary This chapter discusses the disposal of radioactive wastes that arise from a great variety of sources, including the nuclear fuel cycle, beneficial uses of isotopes, and radiation by institutions. Spent fuel contains uranium, plutonium, and highly radioactive fission products. The spent fuel is accumulating, awaiting the development of a high-level waste repository. It is anticipated that a multi-barrier system involving packaging and geologic media will provide protection of the public over the centuries. The favored method of disposal is in a mined cavity deep underground. In some countries, reprocessing the fuel assemblies permits recycling of materials and disposal of smaller volumes of solidified waste. Transportation of wastes is done by casks and containers designed to withstand severe accidents. Low-level wastes come from research and medical procedures and from a variety of activation and fission sources at a reactor site. They generally can be given near-surface burial. Isotopes of special interest are cobalt-60 and cesium-137. Transuranic wastes are being disposed of in the Waste Isolation Pilot Plant. Decommissioning of reactors in the future will contribute a great deal of low-level radioactive waste.

Raymond L. Murray

2001-01-01T23:59:59.000Z

306

X-RAY FLUORESCENCE ANALYSIS OF HANFORD LOW ACTIVITY WASTE SIMULANTS  

SciTech Connect (OSTI)

Savannah River National Laboratory (SRNL) was requested to develop an x-ray fluorescence (XRF) spectrometry method for elemental characterization of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) pretreated low activity waste (LAW) stream to the LAW Vitrification Plant. The WTP is evaluating the potential for using XRF as a rapid turnaround technique to support LAW product compliance and glass former batching. The overall objective of this task was to develop XRF analytical methods that provide the rapid turnaround time (<8 hours) requested by the WTP, while providing sufficient accuracy and precision to determine waste composition variations. For Phase 1a, SRNL (1) evaluated, selected, and procured an XRF instrument for WTP installation, (2) investigated three XRF sample methods for preparing the LAW sub-sample for XRF analysis, and (3) initiated scoping studies on AN-105 (Envelope A) simulant to determine the instrument's capability, limitations, and optimum operating parameters. After preliminary method development on simulants and the completion of Phase 1a activities, SRNL received approval from WTP to begin Phase 1b activities with the objective of optimizing the XRF methodology. Three XRF sample methods used for preparing the LAW sub-sample for XRF analysis were studied: direct liquid analysis, dried spot, and fused glass. The direct liquid method was selected because its major advantage is that the LAW can be analyzed directly without any sample alteration that could bias the method accuracy. It also is the fastest preparation technique--a typical XRF measurement could be completed in < 1hr after sample delivery. Except for sodium, the method detection limits (MDLs) for the most important analytes in solution, the hold point elements, were achieved by this method. The XRF detection limits are generally adequate for glass former batching and product composition reporting, but may be inadequate for some species (Hg, Cd, and Ba) important to land disposal restrictions. The long term precision (24-hr) also was good with percent relative standard deviations (%RSDs) < 10 % for most elements in filtered solution. There were some issues with a few elements precipitating out of solution over time affecting the long term precision of the method. Additional research will need to be performed to resolve this sample stability problem. Activities related to methodology optimization in the Phase 1b portion of the study were eliminated as a result of WTP request to discontinue remaining activities due to funding reduction. These preliminary studies demonstrate that developing an XRF method to support the LAW vitrification plant is feasible. When funding is restored for the WTP, it is recommended that optimization of this technology should be pursued.

Jurgensen, A; David Missimer, D; Ronny Rutherford, R

2006-05-08T23:59:59.000Z

307

Activities for Engaging High School Students in Energy Studies  

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

PARCI-CARES Harry Bolson, LEED Green Associate Washington University in St. Louis PARCI-CARES Activities for Engaging High School Students in Energy Studies Activities for...

308

FLUIDIZED BED STEAM REFORMING MINERALIZATION FOR HIGH ORGANIC AND NITRATE WASTE STREAMS FOR THE GLOBAL NUCLEAR ENERGY PARTNERSHIP  

SciTech Connect (OSTI)

Waste streams that may be generated by the Global Nuclear Energy Partnership (GNEP) Advanced Energy Initiative may contain significant quantities of organics (0-53 wt%) and/or nitrates (0-56 wt%). Decomposition of high nitrate streams requires reducing conditions, e.g. organic additives such as sugar or coal, to reduce the NO{sub x} in the off-gas to N{sub 2} to meet the Clean Air Act (CAA) standards during processing. Thus, organics will be present during waste form stabilization regardless of which GNEP processes are chosen, e.g. organics in the feed or organics for nitrate destruction. High organic containing wastes cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by preprocessing. Alternative waste stabilization processes such as Fluidized Bed Steam Reforming (FBSR) operate at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). FBSR converts organics to CAA compliant gases, creates no secondary liquid waste streams, and creates a stable mineral waste form that is as durable as glass. For application to the high Cs-137 and Sr-90 containing GNEP waste streams a single phase mineralized Cs-mica phase was made by co-reacting illite clay and GNEP simulated waste. The Cs-mica accommodates up to 30% wt% Cs{sub 2}O and all the GNEP waste species, Ba, Sr, Rb including the Cs-137 transmutation to Ba-137. For reference, the cesium mineral pollucite (CsAlSi{sub 2}O{sub 6}), currently being studied for GNEP applications, can only be fabricated at {ge} 1000 C. Pollucite mineralization creates secondary aqueous waste streams and NO{sub x}. Pollucite is not tolerant of high concentrations of Ba, Sr or Rb and forces the divalent species into different mineral host phases. The pollucite can accommodate up to 33% wt% Cs{sub 2}O.

Jantzen, C; Michael Williams, M

2008-01-11T23:59:59.000Z

309

Enhancement of sludge reduction and methane production by removing extracellular polymeric substances from waste activated sludge  

Science Journals Connector (OSTI)

Abstract The management of waste activated sludge (WAS) recycling is a concern that affects the development of the future low-carbon society, particularly sludge reduction and biomass utilization. In this study, we investigated the effect of removing extracellular polymeric substances (EPS), which play important roles in the adhesion and flocculation of WAS, on increased sludge disintegration, thereby enhancing sludge reduction and methane production by anaerobic digestion. EPS removal from WAS by ethylenediaminetetraacetic acid (EDTA) significantly enhanced sludge reduction, i.e., 49 ± 5% compared with 27 ± 1% of the control at the end the digestion process. Methane production was also improved in WAS without EPS by 8881 ± 109 CH4 ?mol g?1 dry-weight of sludge. Microbial activity was determined by denaturing gradient gel electrophoresis and real-time polymerase chain reaction, which showed that the hydrolysis and acetogenesis stages were enhanced by pretreatment with 2% EDTA, with a larger methanogenic community and better methane production.

Minh Tuan Nguyen; Nazlina Haiza Mohd Yasin; Toshiki Miyazaki; Toshinari Maeda

2014-01-01T23:59:59.000Z

310

Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository  

SciTech Connect (OSTI)

A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables.

McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

1983-11-01T23:59:59.000Z

311

Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography  

SciTech Connect (OSTI)

This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

Larson, D.E.

1996-09-01T23:59:59.000Z

312

TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL  

SciTech Connect (OSTI)

5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-07-09T23:59:59.000Z

313

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

SciTech Connect (OSTI)

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

314

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

SciTech Connect (OSTI)

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

315

Radioactive Waste Management (Minnesota)  

Broader source: Energy.gov [DOE]

This section regulates the transportation and disposal of high-level radioactive waste in Minnesota, and establishes a Nuclear Waste Council to monitor the federal high-level radioactive waste...

316

Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste  

Science Journals Connector (OSTI)

Abstract This paper summarizes various hazards identified between 1978 when Yucca Mountain, located in arid southern Nevada, was first proposed as a potential site and 2008 when the license application to construct a repository for spent nuclear fuel and high-level radioactive waste was submitted. Although advantages of an arid site are many, hazard identification and scenario development have generally recognized fractures in the tuff as important features; climate change, water infiltration and percolation, and an oxidizing environment as important processes; and igneous activity, seismicity, human intrusion, and criticality as important disruptive events to consider at Yucca Mountain. Some of the scientific and technical challenges encountered included a change in the repository design from in-floor emplacement with small packages to in-drift emplacement with large packages without backfill. This change, in turn, increased the importance of igneous and seismic hazards.

Rob P. Rechard; Geoff A. Freeze; Frank V. Perry

2014-01-01T23:59:59.000Z

317

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices  

SciTech Connect (OSTI)

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

318

MAR ASSESSMENTS OF THE HIGH LEVEL WASTE SYSTEM PLAN REVISION 16  

SciTech Connect (OSTI)

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit of time) is primarily a function of two critical parameters: waste loading (WL) and melt rate. For the Defense Waste Processing Facility (DWPF), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). Significant increases in waste throughput have been achieved at DWPF since initial radioactive operations began in 1996. Key technical and operational initiatives that supported increased waste throughput included improvements in facility attainment, the Chemical Processing Cell (CPC) flowsheet, process control models and frit formulations. As a result of these key initiatives, DWPF increased WLs from a nominal 28% for Sludge Batch 2 (SB2) to {approx}34 to 38% for SB3 through SB6 while maintaining or slightly improving canister fill times. Although considerable improvements in waste throughput have been obtained, future contractual waste loading targets are nominally 40%, while canister production rates are also expected to increase (to a rate of 325 to 400 canisters per year). Although implementation of bubblers have made a positive impact on increasing melt rate for recent sludge batches targeting WLs in the mid30s, higher WLs will ultimately make the feeds to DWPF more challenging to process. Savannah River Remediation (SRR) recently requested the Savannah River National Laboratory (SRNL) to perform a paper study assessment using future sludge projections to evaluate whether the current Process Composition Control System (PCCS) algorithms would provide projected operating windows to allow future contractual WL targets to be met. More specifically, the objective of this study was to evaluate future sludge batch projections (based on Revision 16 of the HLW Systems Plan) with respect to projected operating windows using current PCCS models and associated constraints. Based on the assessments, the waste loading interval over which a glass system (i.e., a projected sludge composition with a candidate frit) is predicted to be acceptable can be defined (i.e., the projected operating window) which will provide insight into the ability to meet future contractual WL obligations. In this study, future contractual WL obligations are assumed to be 40%, which is the goal after all flowsheet enhancements have been implemented to support DWPF operations. For a system to be considered acceptable, candidate frits must be identified that provide access to at least 40% WL while accounting for potential variation in the sludge resulting from differences in batch-to-batch transfers into the Sludge Receipt and Adjustment Tank (SRAT) and/or analytical uncertainties. In more general terms, this study will assess whether or not the current glass formulation strategy (based on the use of the Nominal and Variation Stage assessments) and current PCCS models will allow access to compositional regions required to targeted higher WLs for future operations. Some of the key questions to be considered in this study include: (1) If higher WLs are attainable with current process control models, are the models valid in these compositional regions? If the higher WL glass regions are outside current model development or validation ranges, is there existing data that could be used to demonstrate model applicability (or lack thereof)? If not, experimental data may be required to revise current models or serve as validation data with the existing models. (2) Are there compositional trends in frit space that are required by the PCCS models to obtain access to these higher WLs? If so, are there potential issues with the compositions of the associated frits (e.g., limitations on the B{sub 2}O{sub 3} and/or Li{sub 2}O concentrations) as they are compared to model development/validation ranges or to the term 'borosilicate' glass? If limitations on the frit compositional range are realized, what is the impact of these restrictions on other glass properties such as the ability to suppress nepheline formation or influence m

Peeler, D.; Edwards, T.

2011-08-05T23:59:59.000Z

319

AIChE issues nuclear waste conclusions  

Science Journals Connector (OSTI)

AIChE issues nuclear waste conclusions ... Long and extensive research, development, and demonstration activities in the U.S. and abroad have made possible the safe, permanent disposal of high-level commercial nuclear waste. ... Titled "The U.S. High-Level Nuclear Waste Management Program," it was prepared by an AIChE nuclear engineering division task force for the institute's government programs steering committee. ...

1986-02-10T23:59:59.000Z

320

Characteristics of potential repository wastes. Volume 3, Appendix 3A, ORIGEN2 decay tables for immobilized high-level waste; Appendix 3B, Interim high-level waste forms  

SciTech Connect (OSTI)

This appendix presents the results of decay calculations using the ORIGEN2 code to determine the radiological properties of canisters of immobilized high-level waste as a function of decay time for decay times up to one million years. These calculations were made for the four HLW sites (West Valley Demonstration Project, Savannah River Site, Hanford Site, and Idaho National Engineering Laboratory) using the composition data discussed in the HLW section of this report. Calculated ({alpha},n) neutron production rates are also shown.

Not Available

1992-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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
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321

THE RETRIEVAL KNOWLEDGE CENTER EVALUATION OF LOW TANK LEVEL MIXING TECHNOLOGIES FOR DOE HIGH LEVEL WASTE TANK RETRIEVAL 10516  

SciTech Connect (OSTI)

The Department of Energy (DOE) Complex has over two-hundred underground storage tanks containing over 80-million gallons of legacy waste from the production of nuclear weapons. The majority of the waste is located at four major sites across the nation and is planned for treatment over a period of almost forty years. The DOE Office of Technology Innovation & Development within the Office of Environmental Management (DOE-EM) sponsors technology research and development programs to support processing advancements and technology maturation designed to improve the costs and schedule for disposal of the waste and closure of the tanks. Within the waste processing focus area are numerous technical initiatives which included the development of a suite of waste removal technologies to address the need for proven equipment and techniques to remove high level radioactive wastes from the waste tanks that are now over fifty years old. In an effort to enhance the efficiency of waste retrieval operations, the DOE-EM Office of Technology Innovation & Development funded an effort to improve communications and information sharing between the DOE's major waste tank locations as it relates to retrieval. The task, dubbed the Retrieval Knowledge Center (RKC) was co-lead by the Savannah River National Laboratory (SRNL) and the Pacific Northwest National Laboratory (PNNL) with core team members representing the Oak Ridge and Idaho sites, as well as, site contractors responsible for waste tank operations. One of the greatest challenges to the processing and closure of many of the tanks is complete removal of all tank contents. Sizeable challenges exist for retrieving waste from High Level Waste (HLW) tanks; with complications that are not normally found with tank retrieval in commercial applications. Technologies currently in use for waste retrieval are generally adequate for bulk removal; however, removal of tank heels, the materials settled in the bottom of the tank, using the same technology have proven to be difficult. Through the RKC, DOE-EM funded an evaluation of adaptable commercial technologies that could assist with the removal of the tank heels. This paper will discuss the efforts and results of developing the RKC to improve communications and discussion of tank waste retrieval through a series of meetings designed to identify technical gaps in retrieval technologies at the DOE Hanford and Savannah River Sites. This paper will also describe the results of an evaluation of commercially available technologies for low level mixing as they might apply to HLW tank heel retrievals.

Fellinger, A.

2009-12-08T23:59:59.000Z

322

SRS - Programs - Liquid Waste Disposition  

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

Liquid Waste Disposition Liquid Waste Disposition This includes both the solidification of highly radioactive liquid wastes stored in SRS's tank farms and disposal of liquid low-level waste generated as a by-product of the separations process and tank farm operations. This low-level waste is treated in the Effluent Treatment Facility. High-activity liquid waste is generated at SRS as by-products from the processing of nuclear materials for national defense, research and medical programs. The waste, totaling about 36 million gallons, is currently stored in 49 underground carbon-steel waste tanks grouped into two "tank farms" at SRS. While the waste is stored in the tanks, it separates into two parts: a sludge that settles on the bottom of the tank, and a liquid supernate that resides on top of the sludge. The waste is reduced to about 30 percent of its original volume by evaporation. The condensed evaporator "overheads" are transferred to the Effluent Treatment Project for final cleanup prior to release to the environment. As the concentrate cools a portion of it crystallizes forming solid saltcake. The concentrated supernate and saltcake are less mobile and therefore less likely to escape to the environment in the event of a tank crack or leak.

323

Chapter 22 - Radioactive Waste Disposal  

Science Journals Connector (OSTI)

Publisher Summary This chapter discusses safe disposal of radioactive waste in order to provide safety to workers and the public. Radioactive wastes arise from a great variety of sources, including the nuclear fuel cycle, and from beneficial uses of isotopes and radiation by institutions. Spent fuel contains uranium, plutonium, and highly radioactive fission products. In the United States spent fuel is accumulating, awaiting the development of a high-level waste repository. A multi-barrier system involving packaging and geological media will provide protection of the public over the centuries the waste must be isolated. The favored method of disposal is in a mined cavity deep underground. In other countries, reprocessing the fuel assemblies permits recycling of materials and disposal of smaller volumes of solidified waste. Transportation of wastes is by casks and containers designed to withstand severe accidents. Low-level wastes (LLWs) come from research and medical procedures and from a variety of activation and fission sources at a reactor site. They generally can be given near-surface burial. Isotopes of special interest are cobalt-60 and cesium-137. Transuranic wastes are being disposed of in the Waste Isolation Pilot Plant. Establishment of regional disposal sites by interstate compacts has generally been unsuccessful in the United States. Decontamination of defense sites will be long and costly. Decommissioning of reactors in the future will contribute a great deal of low-level radioactive waste.

Raymond L. Murray

2009-01-01T23:59:59.000Z

324

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

I I I Y DoE/Els-o120 Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina Volume 2 Q ~<$c'% ~ v ~ g ;:: # +4 -~ STATES O* December 1987 United States Department of Energy -- TABLE OF CONTENTS Appendix A GEOLOGY AND SUBSURFACE HYDROLOGY . . . . . . . . . . . . . . . A.1 Geology and Seismology . . . . . . . . . . . . . . . . . A.1.l Regional Geologic Setting . . . . . . . . . . . . A.1.1.1 Tectonic Provinces . . . . . . . . . . . A.I.1.2 Stratigraphy . . . . . . . . . . . . . . A.1.1.3 Geomorphology . . . . . . . . . . . . . . A.1.2 Seismology and Geologic Hazards . . . . . . . . . A.1.2.1 Geologic Structures and Seismicity . . . A.1.2.2 Seismic Events and Liquefaction Potentill . . . . . . . . . . . . . . . . A.2 Groundwater Resources . . . . . . . . . . . . . . . . . . A.2.1 Hydrostratigraphy . . . . . . . . . . . . . . . . A.2.2 Groundwater Hydrology . . . . . . . . . . . . . . A.2.2.1 Hydrologic Properties

325

Final Environmental Impact Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina  

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

Impact Impact \ DoE/Els-o120 Statement Waste Management Activities for Groundwater Protection Savannah River Plant Aiken, South Carolina Volume 3 Q ~+ ~ FNT O&@+@ &v a w ~ k ~ ;%." $ +6 & ~+e $TiTES Of December 1987 United States Department of Energy TABLE OF CONTENTS ~pendix G ASSESSMENT OF ALTERNATIVE STRATEGIES FOR STORAGE FACILITIES . . . . . . . . . . . G.1 No-Action Strategy . . . . . . . . G.1.l Sununarv and Objectives . . NEW DISPOSAL/ . . . . . . . . . . . . . . . G.1.2 Groundwater and Surface Water Effects G.1.3 Nonradioactive Atmospheric Releases . G.1.4 Ecological Effects . . . . . . . . . G.1.5 Radiological Releases . . . . . . . . G.1.6 Archaeological and Historic Resources G.1.7 SOciOecOnOmics . . . . . . . . . . . G.1.8 Dedication of Site . . . . . . . . . G.1.9 Institutional Impacts . . . . . . . . G.l.10 Noise . . . . . . . . . . . . . . . . G.2 Dedication Strategy . . . . . . . . . . . . . G.2.1 G.2.2 G.2.3 G.2.4 G.2.5 G.2.6

326

Overview of environmental surveillance of waste management activities at the Idaho National Engineering Laboratory  

SciTech Connect (OSTI)

The Idaho National Engineering Laboratory (INEL), in southeastern Idaho, is a principal center for nuclear energy development for the Department of Energy (DOE) and the US Nuclear Navy. Fifty-two reactors have been built at the INEL, with 15 still operable. Extensive environmental surveillance is conducted at the INEL by DOE's Radiological and Environmental Sciences Laboratory (RESL), the US Geological Survey (USGS), the National Oceanic and Atmospheric Administration (NOAA), EG and G Idaho, Inc., and Westinghouse Idaho Nuclear Company (WINCO). Surveillance of waste management facilities is integrated with the overall INEL Site surveillance program. Air, water, soil, biota, and environmental radiation are monitored or sampled routinely at the INEL. Results to date indicate very small or no impacts from the INEL on the surrounding environment. Environmental surveillance activities are currently underway to address key environmental issues at the INEL. 7 refs., 6 figs., 2 tabs.

Smith, T.H.; Hedahl, T.G.; Wiersma, G.B.; Chew, E.W.; Mann, L.J.; Pointer, T.F.

1986-02-01T23:59:59.000Z

327

Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)  

SciTech Connect (OSTI)

Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (K{sub d}) and the thermodynamic solubility product (K{sub sp}), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified construct, called the solution concentration limit, a constant value. In future geochemical data packages, they will determine whether a more rigorous measure of solubility is necessary or warranted based on the dose predictions emanating from the ILAW 2001 PA and reviewers' comments. The K{sub d}s and solution concentration limits for each contaminant are direct inputs to subsurface flow and transport codes used to predict the performance of the ILAW system. In addition to the best-estimate K{sub d}s, a reasonable conservative value and a range are provided. They assume that K{sub d} values are log normally distributed over the cited ranges. Currently, they do not give estimates for the range in solubility limits or their uncertainty. However, they supply different values for both the K{sub d}s and solution concentration limits for different spatial zones in the ILAW system and supply time-varying K{sub d}s for the concrete zone, should the final repository design include concrete vaults or cement amendments to buffer the system pH.

DI Kaplan; RJ Serne

2000-02-24T23:59:59.000Z

328

HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT  

SciTech Connect (OSTI)

The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple intraarea transfers utilizing STPs from July 2006 to August 2007. This operation and successful removal of sludge material meets requirement of approximately 19,000 to 28,000 liters (5,000 to 7,500 gallons) remaining prior to the Chemical Cleaning process. Removal of the last 35% of sludge was exponentially more difficult, as less and less sludge was available to mobilize and the lighter sludge particles were likely removed during the early mixing campaigns. The removal of the 72,000 liters (19,000 gallons) of sludge was challenging due to a number factors. One primary factor was the complex internal cooling coil array within Tank 6 that obstructed mixer discharge jets and impacted the Effective Cleaning Radius (ECR) of the Submersible Mixer Pumps. Minimal access locations into the tank through tank openings (risers) presented a challenge because the available options for equipment locations were very limited. Mechanical Sludge Removal activities using SMPs caused the sludge to migrate to areas of the tank that were outside of the SMP ECR. Various SMP operational strategies were used to address the challenge of moving sludge from remote areas of the tank to the transfer pump. This paper describes in detail the Mechanical Sludge Removal activities and mitigative solutions to cooling coil obstructions and other challenges. The performance of the WOW system and SMP operational strategies were evaluated and the resulting lessons learned are described for application to future Mechanical Sludge Removal operations.

Jolly, R; Bruce Martin, B

2008-01-15T23:59:59.000Z

329

Sequential separation of actinide elements from highly radioactive Hanford waste by ion exchange methods  

SciTech Connect (OSTI)

A simple, rapid method has been developed for the sequential separation of actinide elements from samples with high salt content such as these resulting from efforts to characterize Hanford storage tank waste. Actinide elements in 9M HC1 solution are introduced into an anion exchange column. U(VI), Np(IV) and Pu(IV) are retained on the column while Am(III) passes through. Plutonium is eluted first, reductively; after which neptunium and then uranium are eluted with mixtures of HC1 and HF. The Am(III) is purified by cation exchange in a nitric acid system. 14 refs., 2 tabs.

Maiti, T.C.; Kaye, J.H.; Kozelisky, A.E.

1991-04-01T23:59:59.000Z

330

Performance assessment overview for subseabed disposal of high level radioactive waste  

SciTech Connect (OSTI)

The Subseabed Disposal Project (SDP) was part of an international program that investigated the feasibility of high-level radioactive waste disposal in the deep ocean sediments. This report briefly describes the seven-step iterative performance assessment procedures used in this study and presents representative results of the last iteration. The results of the performance are compared to interim standards developed for the SDP, to other conceptual repositories, and to related metrics. The attributes, limitations, uncertainties, and remaining tasks in the SDP feasibility phase are discussed.

Klett, R.D.

1997-06-01T23:59:59.000Z

331

RADIOACTIVE HIGH LEVEL WASTE TANK PITTING PREDICTIONS: AN INVESTIGATION INTO CRITICAL SOLUTION CONCENTRATIONS  

SciTech Connect (OSTI)

A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel?s susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-level wastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

Hoffman, E.

2012-11-08T23:59:59.000Z

332

Tank Waste Corporate Board Meeting 11/06/08 | Department of Energy  

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

11/06/08 11/06/08 Tank Waste Corporate Board Meeting 11/06/08 The following documents are associated with the Tank Waste Corporate Board Meeting held on November 6th, 2008. Note: (Please contact Steven Ross at steven.ross@em.doe.gov for a HLW Glass Waste Loadings version with animations on slide 6). Slurry Retrieval, Pipeline Transport & Plugging and Mixing Workshop The Way Ahead - West Valley Demonstration Project High-Level Liquid Waste Tank Integrity Workshop - 2008 Savannah River Tank Waste Residuals Hanford Tank Waste Residuals HLW Glass Waste Loadings High-Level Waste Corporate Board Performance Assessment Subcommittee More Documents & Publications Tank Waste Corporate Board Meeting 11/18/10 System Planning for Low-Activity Waste at Hanford Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

333

Waste package degradation from thermal and chemical processes in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste  

Science Journals Connector (OSTI)

Abstract This paper summarizes modeling of waste container degradation in performance assessments conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. As understanding of the Yucca Mountain disposal system increased, modeling of container degradation evolved from a component of the source term in 1984 to a separate module describing both container and drip shield degradation in 2008. A thermal module for evaluating the influence of higher heat loads from more closely packed, large waste packages was also introduced. In addition, a module for evaluating drift chemistry was added in later \\{PAs\\} to evaluate the potential for localized corrosion of the outer barrier of the waste container composed of Alloy 22, a highly corrosion-resistant nickel–chromium–tungsten–molybdenum alloy. The uncertainty of parameters related to container degradation contributed significantly to the estimated uncertainty of performance measures (cumulative release in assessments prior to 1995 and individual dose, thereafter).

Rob P. Rechard; Joon H. Lee; Ernest L. Hardin; Charles R. Bryan

2014-01-01T23:59:59.000Z

334

Evolution of repository and waste package designs for Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste  

Science Journals Connector (OSTI)

Abstract This paper summarizes the evolution of the engineered barrier design for the proposed Yucca Mountain disposal system. Initially, the underground facility used a fairly standard panel and drift layout excavated mostly by drilling and blasting. By 1993, the layout of the underground facility was changed to accommodate construction by a tunnel boring machine. Placement of the repository in unsaturated zone permitted an extended period without backfilling; placement of the waste package in an open drift permitted use of much larger, and thus hotter packages. Hence in 1994, the underground facility design switched from floor emplacement of waste in small, single walled stainless steel or nickel alloy containers to in-drift emplacement of waste in large, double-walled containers. By 2000, the outer layer was a high nickel alloy for corrosion resistance and the inner layer was stainless steel for structural strength. Use of large packages facilitated receipt and disposal of high volumes of spent nuclear fuel. In addition, in-drift package placement saved excavation costs. Options considered for in-drift emplacement included different heat loads and use of backfill. To avoid dripping on the package during the thermal period and the possibility of localized corrosion, titanium drip shields were added for the disposal drifts by 2000. In addition, a handling canister, sealed at the reactor to eliminate further handling of bare fuel assemblies, was evaluated and eventually adopted in 2006. Finally, staged development of the underground layout was adopted to more readily adjust to changes in waste forms and Congressional funding.

Rob P. Rechard; Michael D. Voegele

2014-01-01T23:59:59.000Z

335

Study on the Application of High Temperature Heat Pump to Recover Waste Heat of Marine Diesel Engine  

Science Journals Connector (OSTI)

Being an energy-saving equipment with great development potential, high temperature heat pump is becoming one of the research hotspots in recent years. However, there is little research about the application of high temperature heat pump on ships as ... Keywords: marine diesel engine, cooling water, waste heat recovery, high temperature heat pump

Shi-jie Liu; Wu Chen; Zhen-xiong Cai; Chao-yu Zheng

2009-10-01T23:59:59.000Z

336

Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project  

SciTech Connect (OSTI)

This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B&S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs.

Russell, E.W.; Clarke, W. [Lawrence Livermore National Lab., CA (United States)] [Lawrence Livermore National Lab., CA (United States); Domian, H.A. [Babcock and Wilcox Co., Lynchburg, VA (United States)] [Babcock and Wilcox Co., Lynchburg, VA (United States); Madson, A.A. [Kaiser Engineers California Corp., Oakland, CA (United States)] [Kaiser Engineers California Corp., Oakland, CA (United States)

1991-08-01T23:59:59.000Z

337

Potential role of ABC-assisted repositories in U.S. plutonium and high-level waste disposition  

SciTech Connect (OSTI)

This paper characterizes the issues involving deep geologic disposal of LWR spent fuel rods, then presents results of an investigation to quantify the potential role of Accelerator-Based Conversion (ABC) in an integrated national nuclear materials and high level waste disposition strategy. The investigation used the deep geological repository envisioned for Yucca Mt., Nevada as a baseline and considered complementary roles for integrated ABC transmutation systems. The results indicate that although a U.S. geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of plutonium, the minor actinides, and selected long-lived fission products can result in the following substantial benefits: plutonium burndown to near zero levels, a dramatic reduction of the long term hazard associated with geologic repositories, an ability to place several-fold more high level nuclear waste in a single repository, electricity sales to compensate for capital and operating costs.

Berwald, David; Favale, Anthony; Myers, Timothy; McDaniel, Jerry [Grumman Aerospace Corporation, Bethpage New York 11714 (United States); Bechtel Corporation, 50 Beal St., San Francisco, California 94105 (United States)

1995-09-15T23:59:59.000Z

338

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results  

SciTech Connect (OSTI)

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

339

Legacy Risk Measure for Environmental Management Waste  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating the development of a comprehensive and quantitative risk model framework for environmental management activities at the site. Included are waste management programs (high-level waste, transuranic waste, low-level waste, mixed low-level waste, spent nuclear fuel, and special nuclear materials), major environmental restoration efforts, major decontamination and decommissioning projects, and planned long-term stewardship activities. Two basic types of risk estimates are included: risks from environmental management activities, and long-term legacy risks from wastes/materials. Both types of risks are estimated using the Environment, Safety, and Health Risk Assessment Program (ESHRAP) developed at the INEEL. Given these two types of risk calculations, the following evaluations can be performed: • Risk evaluation of an entire program (covering waste/material as it now exists through disposal or other end states) • Risk comparisons of alternative programs or activities • Comparisons of risk benefit versus risk cost for activities or entire programs • Ranking of programs or activities by risk • Ranking of wastes/materials by risk • Evaluation of site risk changes with time as activities progress • Integrated performance measurement using indicators such as injury/death and exposure rates. This paper discusses the definition and calculation of legacy risk measures and associated issues. The legacy risk measure is needed to support three of the seven types of evaluations listed above: comparisons of risk benefit versus risk cost, ranking of wastes/materials by risk, and evaluation of site risk changes with time.

Eide, Steven Arvid; Nitschke, Robert Leon

2002-02-01T23:59:59.000Z

340

High-temperature photochemical destruction of toxic organic wastes using concentrated solar radiation  

SciTech Connect (OSTI)

Application of concentrated solar energy has been proposed to be a viable waste disposal option. Specifically, this concept of solar induced high-temperature photochemistry is based on the synergistic contribution of concentrated infrared (IR) radiation, which acts as an intense heating source, and near ultraviolet and visible (UV-VIS) radiation, which can induce destructive photochemical processes. Some significant advances have been made in the theoretical framework of high-temperature photochemical processes (Section 2) and development of experimental techniques for their study (Section 3). Basic thermal/photolytic studies have addressed the effect of temperature on the photochemical destruction of pure compounds (Section 4). Detailed studies of the destruction of reaction by-products have been conducted on selected waste molecules (Section 5). Some very limited results are available on the destruction of mixtures (Section 6). Fundamental spectroscopic studies have been recently initiated (Section 7). The results to date have been used to conduct some relatively simple scale-up studies of the solar detoxification process. More recent work has focused on destruction of compounds that do not directly absorb solar radiation. Research efforts have focused on homogeneous as well as heterogeneous methods of initiating destructive reaction pathways (Section 9). Although many conclusions at this point must be considered tentative due to lack of basic research, a clearer picture of the overall process is emerging (Section 10). However, much research remains to be performed and most follow several veins, including photochemical, spectroscopic, combustion kinetic, and engineering scale-up (Section 11).

Dellinger, B.; Graham, J.L.; Berman, J.M.; Taylor, P.H. [Dayton Univ., OH (United States)

1994-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07  

SciTech Connect (OSTI)

This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

2013-11-13T23:59:59.000Z

342

Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers  

SciTech Connect (OSTI)

Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

1995-06-01T23:59:59.000Z

343

Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository  

SciTech Connect (OSTI)

This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking, and hydrogen induced cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure.

K.G. Mon; F. Hua

2005-04-12T23:59:59.000Z

344

Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096  

SciTech Connect (OSTI)

Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States)] [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)] [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

2013-07-01T23:59:59.000Z

345

Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 1  

SciTech Connect (OSTI)

The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

Hofmann, P.L.; Breslin, J.J. (eds.)

1981-01-01T23:59:59.000Z

346

Idaho National Engineering Laboratory High-Level Waste Roadmap. Revision 2  

SciTech Connect (OSTI)

The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ``where we are now`` to ``where we want and need to be.`` The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues.

Not Available

1993-08-01T23:59:59.000Z

347

Design and performance of atomizing nozzles for spray calcination of high-level wastes  

SciTech Connect (OSTI)

A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-..mu.. median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-..mu.. median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure.

Miller, F.A.; Stout, L.A.

1981-05-01T23:59:59.000Z

348

MATRIX 1 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL  

SciTech Connect (OSTI)

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). It has been proposed that a team of glass formulation and processing experts at the Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), and Vitreous State Laboratory (VSL) at Catholic University of America develop a systematic approach to increase HLW throughput (by increasing WL with minimal or positive impacts on melt rate). Programmatically, this task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. The following four specific tasks have been proposed to meet this programmatic objective: (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. The details of these tasks can be found in the associated task plan WSRC-STI-2007-00483. The current study is focused on Task 2 (crystal accumulation modeling and higher waste loading glasses) and involves glass formulation and physical property testing by both PNNL and SRNL (as defined in the PNNL and SRNL test plans). The intent of this report is to document the chemical composition and Product Consistency Test (PCT) results and statistical analysis of PNNL's Test Matrix 1 glasses. Note that this document is only a compilation of the data collected by SRNL for PNNL's glasses in support of this task and no conclusions will be drawn.

Raszewski, F; Tommy Edwards, T; David Peeler, D

2008-09-23T23:59:59.000Z

349

Waste Growth Challenges Local Democracy. The Politics of Waste between Europe and the Mediterranean: a Focus on Italy  

E-Print Network [OSTI]

activities, such as waste burning versus waste dumping.and the Geographies of Waste Governance: A Burning Issue forEurope: • Burning oriented – Incineration (waste-to-energy)

Mengozzi, Alessandro

2010-01-01T23:59:59.000Z

350

Siting America's geologic repository for high-level Nuclear Waste: Implications for environmental Policy  

Science Journals Connector (OSTI)

Siting a geologic repository for isolating highlevel nuclear waste up to 10,000 years is ... before attempted in the United States. The Nuclear Waste Policy Act of 1982 exempted repository siting from ... require...

John Lemons; Charles Malone

351

AVS: Experimental Tests of a New Process to Inductively Vitrify HLW Inside the Final Disposal Containers at Very High Waste Loadings  

SciTech Connect (OSTI)

The design and performance capabilities of the Advanced Vitrification System (AVS) are described, together with the results of experimental tests. The AVS is an in-can melting system in which high-level waste (HLW) is vitrified directly inside the final disposal container. The AVS container, or module, consists of an outer stainless steel canister and an alumina-lined, inner graphite crucible, which is thermally insulated from the outer stainless canister. The graphite crucible is inductively heated to very high temperatures (up to 1500 C) by an external low frequency (30 Hertz) alternating current (AC) transformer coil. The actively cooled outer stainless canister remains at near ambient temperature. The HLW/frit mixture is fed into the hot graphite crucible, where it is vitrified. After cooldown, the HLW/frit feed and off-gas pipes are disconnected from the top of the module, which is then sealed and readied for shipment or storage. All radioactively contaminated melter components inside the module are disposed of along with the vitrified waste. The graphite crucible also provides a geologically stable barrier for the vitrified product. The AVS potentially can double HLW loading over that obtained from Joule melters; lower vitrification costs by about half; reduce the number of disposal canisters required by about half; handle diverse waste feeds with high concentrations of problem elements such as chromium and zirconium; and reduce the time needed to vitrify a given inventory of HLW.

Powell, J.; Reich, M.; Jordan, J.; Ventre, L.; Barletta, R.; Manowitz, B.; Steinberg, M.; Grossman, W.; Maise, G.; Salzano, F.; Hess, C.; Ramsey, W. G.; Plodinec, M. J.

2002-02-26T23:59:59.000Z

352

Tank Waste Corporate Board Meeting 11/18/10 | Department of Energy  

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

Tank Waste Corporate Board Meeting 11/18/10 Tank Waste Corporate Board Meeting 11/18/10 Tank Waste Corporate Board Meeting 11/18/10 The following documents are associated with the Tank Waste Corporate Board Meeting held on November 18th, 2010. High-Level Waste Corporate Board Meeting Agenda Journey to Excellence Goal 2 and Enhanced Tank Waste Strategy Introduction to Tc/I in Hanford Flowsheet Fate of Tc99 at WTP and Current Work on Capture Technetium Retention During LAW Vitrification Impacts of Feed Composition and Recycle on Hanford Low-Activity Waste Glass Mass Secondary Waste Forms and Technetium Management Hanford Low Activity Waste (LAW) Fluidized Bed Steam Reformer (FBSR) Na-Al-Si (NAS) Waste Form Qualification Salt Waste Processing Initiatives Recap and Conclusions to Tc/I in Hanford Flowsheet Presentations

353

Independent Oversight Activity Report, Hanford Waste Tank Farms – October 28 – November 6, 2013  

Broader source: Energy.gov [DOE]

Follow-up on Previously Identified Items Regarding Positive Ventilation of Hanford Underground Waste Tanks [HIAR-HANFORD-2013-10-28

354

Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site  

SciTech Connect (OSTI)

processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been observed in any of the pour stream glass samples. Spinel was observed at the bottom of DWPF Melter 1 as a result of K-3 refractory corrosion. Issues have occurred with accumulation of spinel in the pour spout during periods of operation at higher waste loadings. Given that both DWPF melters were or have been in operation for greater than 8 years, the service life of the melters has far exceeded design expectations. It is possible that the DWPF liquidus temperature approach is conservative, in that it may be possible to successfully operate the melter with a small degree of allowable crystallization in the glass. This could be a viable approach to increasing waste loading in the glass assuming that the crystals are suspended in the melt and swept out through the riser and pour spout. Additional study is needed, and development work for WTP might be leveraged to support a different operating limit for the DWPF. Several recommendations are made regarding considerations that need to be included as part of the WTP crystal tolerant strategy based on the DWPF development work and operational data reviewed here. These include: Identify and consider the impacts of potential heat sinks in the WTP melter and glass pouring system; Consider the contributions of refractory corrosion products, which may serve to nucleate additional crystals leading to further accumulation; Consider volatilization of components from the melt (e.g., boron, alkali, halides, etc.) and determine their impacts on glass crystallization behavior; Evaluate the impacts of glass REDuction/OXidation (REDOX) conditions and the distribution of temperature within the WTP melt pool and melter pour chamber on crystal accumulation rate; Consider the impact of precipitated crystals on glass viscosity; Consider the impact of an accumulated crystalline layer on thermal convection currents and bubbler effectiveness within the melt pool; Evaluate the impact of spinel accumulation on Joule heating of the WTP melt pool; and Include noble metals in glass melt experiments because of their potential to act as nucleation site

Fox, K. M.

2014-02-27T23:59:59.000Z

355

PERFORMANCE OF A BURIED RADIOACTIVE HIGH LEVEL WASTE GLASS AFTER 24 YEARS  

SciTech Connect (OSTI)

A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

Jantzen, C; Daniel Kaplan, D; Ned Bibler, N; David Peeler, D; John Plodinec, J

2008-05-05T23:59:59.000Z

356

TWRS retrieval and disposal mission, immobilized high-level waste storage plan  

SciTech Connect (OSTI)

This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan`s dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

Calmus, R.B.

1998-01-07T23:59:59.000Z

357

Vapor Corrosion Response of Low Carbon Steel Exposed to Simulated High Level Radioactive Waste  

SciTech Connect (OSTI)

A program to resolve the issues associated with potential vapor space corrosion and liquid/air interface corrosion in the Type III high level waste tanks is in place. The objective of the program is to develop understanding of vapor space (VSC) and liquid/air interface (LAIC) corrosion to ensure a defensible technical basis to provide accurate corrosion evaluations with regard to vapor space and liquid/air interface corrosion. The results of the FY05 experiments are presented here. The experiments are an extension of the previous research on the corrosion of tank steel exposed to simple solutions to corrosion of the steel when exposed to complex high level waste simulants. The testing suggested that decanting and the consequent residual species on the tank wall is the predominant source of surface chemistry on the tank wall. The laboratory testing has shown that at the boundary conditions of the chemistry control program for solutions greater than 1M NaNO{sub 3}{sup -}. Minor and isolated pitting is possible within crevices in the vapor space of the tanks that contain stagnant dilute solution for an extended period of time, specifically when residues are left on the tank wall during decanting. Liquid/air interfacial corrosion is possible in dilute stagnant solutions, particularly with high concentrations of chloride. The experimental results indicate that Tank 50 would be most susceptible to the potential for liquid/air interfacial corrosion or vapor space corrosion, with Tank 49 and 41 following, since these tanks are nearest to the chemistry control boundary conditions. The testing continues to show that the combination of well-inhibited solutions and mill-scale sufficiently protect against pitting in the Type III tanks.

Wiersma, B

2006-01-26T23:59:59.000Z

358

High Level Overview of DOE Biomass Logistics II Project Activities  

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

High Level Overview of DOE Biomass Logistics II Project Activities Biomass 2014: Growing the Future Bioeconomy Washington, DC July 29, 2014 Presented By: Kevin Comer, Associate...

359

Characterization, Propagation and Analysis of Aleatory and Epistemic Uncertainty in the 2008 Performance Assessment for the Proposed Repository for High-Level Radioactive Waste at Yucca Mountain, Nevada  

Science Journals Connector (OSTI)

The 2008 performance assessment (PA) for the proposed repository for high-level radioactive waste at Yucca Mountain (YM), Nevada, illustrates the conceptual...

Clifford W. Hansen; Jon C. Helton; Cédric J. Sallaberry

2010-01-01T23:59:59.000Z

360

Tank waste treatment R and D activities at Oak Ridge National Laboratory  

SciTech Connect (OSTI)

Oak Ridge National Laboratory (ORNL) served as the pilot plant for the Hanford production facility during the 1940s. As a result, the waste contained in the ORNL storage tanks has similarities to waste found at other sites, but is typically 10 to 100 times less radioactive. It is estimated that nearly 4.9 million liters of legacy of waste is stored on the site of ORNL. Of this volume about one-fifth is transuranic sludges. The remainder of the waste volume is classified as low-level waste. The waste contains approximately 130,000 Ci, composed primarily of {sup 137}Cs, {sup 90}Sr, and small amounts of other fission products. The wastes were originally acidic in nature but were neutralized using Na{sub 2}CO{sub 3}, NaOH, or CaO to allow their storage in tanks constructed of carbon steel or concrete (Gunite). In addition to the legacy waste, about 57,000 L of concentrated waste is generated annually, which contains about 13,000 Ci, consisting primarily of {sup 137}Cs, {sup 90}Sr, and small amounts of other fission products. As part of the US department of Energy`s (DOE`s) Environmental Management Tanks Focus Area and Efficient Separations and Processing programs, a number of tasks are under way at ORNL to address the wastes currently stored in tanks across the DOE complex. This paper summarizes the efforts in three of these tasks: (1) the treatment of the tank supernatant to remove Cs, Tc, and Sr; (2) the leaching or washing of the sludges to reduce the volume of waste to be vitrified; and (3) the immobilization of the sludges.

Jubin, R.T.; Lee, D.D.; Beahm, E.C.; Collins, J.L.; Davidson, D.J.; Egan, B.Z.; Mattus, A.J.; Walker, J.F. Jr. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.

1997-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

Risk assessment for the off-site transportation of high-level waste for the U.S. Department of Energy waste management programmatic environmental impact statement  

SciTech Connect (OSTI)

This report describes the human health risk assessment conducted for the transportation of high-level waste (HLW) in support of the US Department of Energy Waste Management Programmatic Environmental Impact Statement (WM PEIS). The assessment considers risks to collective populations and individuals under both routine and accident transportation conditions for truck and rail shipment modes. The report discusses the scope of the HLW transportation assessment, describes the analytical methods used for the assessment, defines the alternatives considered in the WM PEIS, and details important assessment assumptions. Results are reported for five alternatives. In addition, to aid in the understanding and interpretation of the results, specific areas of uncertainty are described, with an emphasis on how the uncertainties may affect comparisons of the alternatives.

Monette, F.A.; Biwer, B.M.; LePoire, D.J.; Chen, S.Y. [Argonne National Lab., IL (United States). Environmental Assessment Div.

1996-12-01T23:59:59.000Z

362

Nuclear Waste: Knowledge Waste?  

Science Journals Connector (OSTI)

...06520, USA. Nuclear power is re-emerging...proclaiming a “nuclear renaissance...example, plant safety...liabilities, terrorism at plants and in transport...high-level nuclear wastes (HLW...factor in risk perceptions...supporting nuclear power in the abstract...

Eugene A. Rosa; Seth P. Tuler; Baruch Fischhoff; Thomas Webler; Sharon M. Friedman; Richard E. Sclove; Kristin Shrader-Frechette; Mary R. English; Roger E. Kasperson; Robert L. Goble; Thomas M. Leschine; William Freudenburg; Caron Chess; Charles Perrow; Kai Erikson; James F. Short

2010-08-13T23:59:59.000Z

363

Tritium waste disposal technology in the US  

SciTech Connect (OSTI)

Tritium waste disposal methods in the US range from disposal of low specific activity waste along with other low-level waste in shallow land burial facilities, to disposal of kilocurie amounts in specially designed triple containers in 65' deep augered holes located in an aird region of the US. Total estimated curies disposed of are 500,000 in commercial burial sites and 10 million curies in defense related sites. At three disposal sites in humid areas, tritium has migrated into the ground water, and at one arid site tritium vapor has been detected emerging from the soil above the disposal area. Leaching tests on tritium containing waste show that tritium in the form of HTO leaches readily from most waste forms, but that leaching rates of tritiated water into polymer impregnated concrete are reduced by as much as a factor of ten. Tests on improved tritium containment are ongoing. Disposal costs for tritium waste are 7 to 10 dollars per cubic foot for shallow land burial of low specific activity tritium waste, and 10 to 20 dollars per cubic foot for disposal of high specific activity waste. The cost of packaging the high specific activity waste is 150 to 300 dollars per cubic foot. 18 references.

Albenesius, E.L.; Towler, O.A.

1983-01-01T23:59:59.000Z

364

Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission.

Lumetta, Gregg J.; Braley, Jenifer C.; Edwards, Matthew K.; Qafoku, Odeta; Felmy, Andrew R.; Carter, Jennifer C.; MacFarlan, Paul J.

2010-09-22T23:59:59.000Z

365

Assessment of fission product content of high-level liquid waste supernate on E-Area vault package criteria  

SciTech Connect (OSTI)

This report assesses the tank farm`s high level waste supernate to determine any potential impacts on waste certification for the E-Area vaults (EAV). The Waste Acceptance Criteria procedure (i.e., WAC 3.10 of the 1S manual) imposes administrative controls on radioactive material in waste packages sent to the EAV, specifically on six fission products. Waste tank supernates contain various fission products, so any waste package containing material contaminated with supernate will contain these radioactive isotopes. This report develops the process knowledge basis for characterizing the supernate composition for these isotopes, so that appropriate controls can be implemented to ensure that the EAV WAC is met. Six fission products are listed in the SRS 1S Manual WAC 3.10: Se-79, which decays to bromine; Sr-90, which decays to niobium; Tc-99, which decays to ruthenium; Sn-126, which decays to tellurium; I-129, which decays to xenon; and Cs-137, which decays to barium.

Brown, D.F.

1994-06-30T23:59:59.000Z

366

Technologies for destruction of long-lived radionuclides in high-level nuclear waste: Overview and requirements  

SciTech Connect (OSTI)

This paper, and this topical session on Nuclear Waste Minimization, Management and Remediation, focuses on two nuclear systems, and their associated technologies, that have the potential to address concerns surrounding long-lived radionuclides in high-level waste. Both systems offer technology applicable to HLW from present light-water reactors (LWR). Additionally these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes, long-term fuel supplies, and enhanced safety. The first system is the Integral Fast Reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while simultaneously burning long-lived actinide waste. IFR applications include burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, Accelerator Transmutation of Waste (ATW), is based upon an accelerator-induced intense source of thermal neutrons and is aimed at destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction.

Arthur, E.D.

1993-10-01T23:59:59.000Z

367

Waste Treatment Plant - 12508  

SciTech Connect (OSTI)

The Waste Treatment Plant (WTP) will immobilize millions of gallons of Hanford's tank waste into solid glass using a proven technology called vitrification. The vitrification process will turn the waste into a stable glass form that is safe for long-term storage. Our discussion of the WTP will include a description of the ongoing design and construction of this large, complex, first-of-a-kind project. The concept for the operation of the WTP is to separate high-level and low-activity waste fractions, and immobilize those fractions in glass using vitrification. The WTP includes four major nuclear facilities and various support facilities. Waste from the Tank Farms is first pumped to the Pretreatment Facility at the WTP through an underground pipe-in-pipe system. When construction is complete, the Pretreatment Facility will be 12 stories high, 540 feet long and 215 feet wide, making it the largest of the four major nuclear facilities that compose the WTP. The total size of this facility will be more than 490,000 square feet. More than 8.2 million craft hours are required to construct this facility. Currently, the Pretreatment Facility is 51 percent complete. At the Pretreatment Facility the waste is pumped to the interior waste feed receipt vessels. Each of these four vessels is 55-feet tall and has a 375,000 gallon capacity, which makes them the largest vessels inside the Pretreatment Facility. These vessels contain a series of internal pulse-jet mixers to keep incoming waste properly mixed. The vessels are inside the black-cell areas, completely enclosed behind thick steel-laced, high strength concrete walls. The black cells are designed to be maintenance free with no moving parts. Once hot operations commence the black-cell area will be inaccessible. Surrounded by black cells, is the 'hot cell canyon'. The hot cell contains all the moving and replaceable components to remove solids and extract liquids. In this area, there is ultrafiltration equipment, cesium-ion exchange columns, evaporator boilers and recirculation pumps, and various mechanical process pumps for transferring process fluids. During the first phase of pretreatment, the waste will be concentrated using an evaporation process. Solids will be filtered out, and the remaining soluble, highly radioactive isotopes will be removed using an ion-exchange process. The high-level solids will be sent to the High-Level Waste (HLW) Vitrification Facility, and the low activity liquids will be sent to the Low-Activity Waste (LAW) Vitrification Facility for further processing. The high-level waste will be transferred via underground pipes to the HLW Facility from the Pretreatment Facility. The waste first arrives at the wet cell, which rests inside a black-cell area. The pretreated waste is transferred through shielded pipes into a series of melter preparation and feed vessels before reaching the melters. Liquids from various facility processes also return to the wet cell for interim storage before recycling back to the Pretreatment Facility. (authors)

Harp, Benton; Olds, Erik [US DOE (United States)

2012-07-01T23:59:59.000Z

368

The effect of high-level waste glass composition on spinel liquidus temperature  

SciTech Connect (OSTI)

Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Riley, Brian J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hrma, Pavel [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2012-11-15T23:59:59.000Z

369

C-106 High-Level Waste Solids: Washing/Leaching and Solubility Versus Temperature Studies  

SciTech Connect (OSTI)

This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-level waste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

GJ Lumetta; DJ Bates; PK Berry; JP Bramson; LP Darnell; OT Farmer III; LR Greenwood; FV Hoopes; RC Lettau; GF Piepel; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

2000-01-26T23:59:59.000Z

370

Phase chemistry and radionuclide retention of high level radioactive waste tank sludges  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has millions of gallons of high level nuclear waste stored in underground tanks at Hanford, Washington and Savannah River, South Carolina. These tanks will eventually be emptied and decommissioned. This will leave a residue of sludge adhering to the interior tank surfaces that may contaminate groundwaters with radionuclides and RCRA metals. Experimentation on such sludges is both dangerous and prohibitively expensive so there is a great advantage to developing artificial sludges. The US DOE Environmental Management Science Program (EMSP) has funded a program to investigate the feasibility of developing such materials. The following text reports on the success of this program, and suggests that much of the radioisotope inventory left in a tank will not move out into the surrounding environment. Ultimately, such studies may play a significant role in developing safe and cost effective tank closure strategies.

KRUMHANSL,JAMES L.; BRADY,PATRICK V.; ZHANG,PENGCHU; ARTHUR,SARA E.; HUTCHERSON,SHEILA K.; LIU,J.; QIAN,M.; ANDERSON,HOWARD L.

2000-05-19T23:59:59.000Z

371

The effect of high-level waste glass composition on spinel liquidus temperature  

SciTech Connect (OSTI)

Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel solubility (c0) is a function of both glass composition and temperature (T). Previously reported models of TL as a function of composition are based on TL measured directly, which requires laborious experimental procedures. Viewing the curve of c0 versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates TL as a function of composition based on c0 data obtained with the X-ray diffraction technique.

Hrma, Pavel R.; Riley, Brian J.; Crum, Jarrod V.; Matyas, Josef

2014-01-15T23:59:59.000Z

372

Interim radiological safety standards and evaluation procedures for subseabed high-level waste disposal  

SciTech Connect (OSTI)

The Seabed Disposal Project (SDP) was evaluating the technical feasibility of high-level nuclear waste disposal in deep ocean sediments. Working standards were needed for risk assessments, evaluation of alternative designs, sensitivity studies, and conceptual design guidelines. This report completes a three part program to develop radiological standards for the feasibility phase of the SDP. The characteristics of subseabed disposal and how they affect the selection of standards are discussed. General radiological protection standards are reviewed, along with some new methods, and a systematic approach to developing standards is presented. The selected interim radiological standards for the SDP and the reasons for their selection are given. These standards have no legal or regulatory status and will be replaced or modified by regulatory agencies if subseabed disposal is implemented. 56 refs., 29 figs., 15 tabs.

Klett, R.D.

1997-06-01T23:59:59.000Z

373

Cyclic voltammetric studies for the electrochemical determination of palladium in high-level nuclear waste  

Science Journals Connector (OSTI)

Cyclic voltammetric studies of Pd(II)/Pd(0) electrode process were carried out on a Glassy Carbon Electrode (GCE) in HCl for the development of an Anodic Stripping Voltammetric (ASV) method for the determination of palladium in High-Level Nuclear Waste (HLNW). Pd(II) reduces electrochemically to Pd(0) in a wide potential range, depending upon the concentration of HCl. No significant effect of concentration was observed on the oxidation of palladium, which more or less occurs at 500 mV. Effects of HCl concentration, potential scan range, scan rate and scan repetition were studied in detail. The oxidation of palladium in HCl medium was relatively more distinct than in nitric and sulphuric acids. Maximum anodic and cathodic peak currents of unequal heights were observed at 1.0 × 10?2 M concentration of HCl. An ASV method was developed successfully on the basis of these studies for the determination of palladium in HLNW.

T.K. Bhardwaj

2012-01-01T23:59:59.000Z

374

Technologies for destruction of long-lived radionuclides in high-level nuclear waste - overview and requirements  

SciTech Connect (OSTI)

A major issue surrounding current nuclear power generation is the management and disposal of long-lived, high-level waste (HLW). The planned and scientifically acceptable destination for this waste is in deep underground, geologically stable, repositories. However, public concerns surrounding such disposal of long-lived nuclear wastes and other issues such as proliferation and safety negatively affect the potential role that nuclear power can play in meeting current and future national energy needs. This paper and this topical session on nuclear waste minimization, management, and remediation focus on two nuclear systems and their associated technologies that have the potential to address concerns surrounding long-lived radionuclides in HLW. Both systems offer technology applicable to HLW from current light water reactors (LWRs). In addition, these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes long-term fuel supplies, and enhanced safety. The first system is the integral fast reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while burning long-lived actinide waste. The IFR applications include the burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, accelerator transmutation of waste (ATW), is based on an accelerator-induced intense source of thermal neutrons and is aimed at the destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent-fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction.

Arthur, E.D. (Los Alamos National Lab., NM (United States))

1993-01-01T23:59:59.000Z

375

Impacts of Feed Composition and Recycle on Hanford Low-Activity Waste Glass Mass  

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

Feed Composition Feed Composition and Recycle on Hanford Low- Activity Waste Glass Mass J.D. Vienna & D.S. Kim - Pacific Northwest National Laboratory I.L. Pegg - Catholic University of America 1 LAW Glass Loading Limits WTP baseline (LAW Glass Formulation Algorithm) low uncertainty  thoroughly tested accounts for Na, S, Cl, F, Cr, K, and P impacts conservative loading Advanced silicate formulation higher uncertainty than baseline currently accounts for Na and S impacts impacts of other components not specifically tested, but, one can evaluate maxima from testing as a lower bound 2 0 5 10 15 20 25 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 SO 3 (target) wt% in Glass Na 2 O (target) wt% in Glass SO 3 ≤ 0.77 wt% Na 2 O ≤ 35.875 - 42.5*SO 3 (in wt%) Na 2 O ≤ 21 wt% Na 2 O + 0.66*K

376

Crystallization of rhenium salts in a simulated low-activity waste borosilicate glass  

SciTech Connect (OSTI)

This study presents a new method for looking at the solubility of volatile species in simulated low-activity waste glass. The present study looking at rhenium salts is also applicable to real applications involving radioactive technetium salts. In this synthesis method, oxide glass powder is mixed with the volatiles species, vacuum-sealed in a fused quartz ampoule, and then heat-treated under vacuum in a furnace. This technique restricts the volatile species to the headspace above the melt but still within the sealed ampoule, thus maximizing the volatile concentration in contact with the glass. Various techniques were used to measure the solubility of rhenium in glass and include energy dispersive spectroscopy, wavelength dispersive spectroscopy, laser ablation inductively-coupled plasma mass spectroscopy, and inductively-coupled plasma optical emission spectroscopy. The Re-solubility in this glass was determined to be ~3004 parts per million Re atoms. Above this concentration, the salts separated out of the melt as inclusions and as a low viscosity molten salt phase on top of the melt observed during and after cooling. This salt phase was analyzed with X-ray diffraction, scanning electron microscopy as well as some of the other aforementioned techniques and identified to be composed of alkali perrhenate and alkali sulfate.

Riley, Brian J.; McCloy, John S.; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J.; Liu, Juan; Rodriguez, Carmen P.; Kim, Dong-Sang

2013-04-01T23:59:59.000Z

377

Design methodology to develop a conceptual underground facility for the disposal of high-level nuclear waste at Yucca Mountain, Nevada  

SciTech Connect (OSTI)

This paper examines the design methodology employed to develop conceptual underground layouts for a prospective high level nuclear waste repository at Yucca Mountain, Nevada. This study is in conjunction with the Nevada Nuclear Waste Storage Investigations (NNWSI), project studying the disposal of high level waste in densely welded tuff. The fundamental design effort concentraes on the effects of the heat released from the decaying waste forms and the impact of this heat on ventilation, waste emplacement configurations, and rock stability. This effort will perfect the design of the waste emplacement layout including emplacement hole spacing, emplacement drift spacing, and the areal power density (APD) for the installed waste. This paper contains only viewgraphs. 11 figs.

Zerga, D.P.; Badie, A.

1986-12-31T23:59:59.000Z

378

Pilot-scale anaerobic co-digestion of municipal biomass waste and waste activated sludge in China: Effect of organic loading rate  

SciTech Connect (OSTI)

Highlights: Black-Right-Pointing-Pointer Co-digestion of municipal biomass waste (MBW) and waste activated sludge (WAS) was examined on a pilot-scale reactor. Black-Right-Pointing-Pointer System performance and stability under OLR of 1.2, 2.4, 3.6, 4.8, 6.0 and 8.0 kg VS (m{sup 3} d){sup -1} were analyzed. Black-Right-Pointing-Pointer A maximum methane production rate of 2.94 m{sup 3} (m{sup 3} d){sup -1} was achieved at OLR of 8.0 kg VS (m{sup 3} d){sup -1} and HRT of 15d. Black-Right-Pointing-Pointer With the increasing OLRs, pH values, VS removal rate and methane concentration decreased and VFA increased. Black-Right-Pointing-Pointer The changing of biogas production rate can be a practical approach to monitor and control anaerobic digestion system. - Abstract: The effects of organic loading rate on the performance and stability of anaerobic co-digestion of municipal biomass waste (MBW) and waste activated sludge (WAS) were investigated on a pilot-scale reactor. The results showed that stable operation was achieved with organic loading rates (OLR) of 1.2-8.0 kg volatile solid (VS) (m{sup 3} d){sup -1}, with VS reduction rates of 61.7-69.9%, and volumetric biogas production of 0.89-5.28 m{sup 3} (m{sup 3} d){sup -1}. A maximum methane production rate of 2.94 m{sup 3} (m{sup 3} d){sup -1} was achieved at OLR of 8.0 kg VS (m{sup 3} d){sup -1} and hydraulic retention time of 15 days. With increasing OLRs, the anaerobic reactor showed a decrease in VS removal rate, average pH value and methane concentration, and a increase of volatile fatty acid concentration. By monitoring the biogas production rate (BPR), the anaerobic digestion system has a higher acidification risk under an OLR of 8.0 kg VS (m{sup 3} d){sup -1}. This result remarks the possibility of relating bioreactor performance with BPR in order to better understand and monitor anaerobic digestion process.

Liu Xiao, E-mail: liuxiao07@mails.tsinghua.edu.cn [School of Environment, Tsinghua University, Beijing 100084 (China); Wang Wei; Shi Yunchun; Zheng Lei [School of Environment, Tsinghua University, Beijing 100084 (China); Gao Xingbao [Chinese Research Academy of Environmental Sciences, Beijing 100012 (China); Qiao Wei [State Key Laboratory of Heavy Oil Processing, China University of Petroleum, Beijing 102249 (China); Zhou Yingjun [Department of Urban and Environmental Engineering, Graduate School of Engineering, Kyoto University, Katsura, Nisikyo-ku, Kyoto 615-8540 (Japan)

2012-11-15T23:59:59.000Z

379

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests  

SciTech Connect (OSTI)

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

2013-01-17T23:59:59.000Z

380

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342  

SciTech Connect (OSTI)

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States)] [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)] [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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

INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS  

SciTech Connect (OSTI)

The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted following an analytical plan. A review of the individual oxides for each glass revealed that there were no errors in batching significant enough to impact the outcome of the study. A comparison of the measured compositions of the replicates indicated an acceptable degree of repeatability as the percent differences for most of the oxides were less than 5% and percent differences for all of the oxides were less than 10 wt%. Chemical durability was measured using the Product Consistency Test (PCT). All but two of the study glasses had normalized leachate for boron (NL [B]) values that were well below that of the Environmental Assessment (EA) reference glass. The two highest NL [B] values were for the CCC versions of glasses US-18 and US-27 (10.498 g/L and 15.962 g/L, respectively). Nepheline crystallization was identified by qualitative XRD in five of the US-series glasses. Each of these five glasses (US-18, US-26, US-27, US-37 and US-43) showed a significant increase in NL [B] values after the CCC heat treatment. This reduction in durability can be attributed to the formation of nepheline during the slow cooling cycle and the removal of glass formers from the residual glass network. The liquidus temperature (T{sub L}) of each glass in the study was determined by both optical microscopy and XRD methods. The correlation coefficient of the measured XRD TL data versus the measured optical TL data was very good (R{sup 2} = 0.9469). Aside from a few outliers, the two datasets aligned very well across the entire temperature range (829 C to 1312 C for optical data and 813 C to 1310 C for XRD crystal fraction data). The data also correlated well with the predictions of a PNNL T{sub L} model. The correlation between the measured and calculated data had a higher degree of merit for the XRD crystal fraction data than for the optical data (higher R{sup 2} value of 0.9089 versus 0.8970 for the optical data). The SEM-EDS analysis of select samples revealed the presence of undissolved RuO{sub 2} in all glasses due to the low solubility of RuO{sub 2} in borosilicate glass. These

Fox, K; David Peeler, D; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P; James Marra, J

2008-09-23T23:59:59.000Z

382

Waste Form Development for the Solidification of PDCF/MOX Liquid Waste Streams  

SciTech Connect (OSTI)

At the Savannah River Site, part of the Department of Energy's nuclear materials complex located in South Carolina, cementation has been selected as the solidification method for high-alpha and low-activity waste streams generated in the planned plutonium disposition facilities. A Waste Solidification Building (WSB) that will be used to treat and solidify three radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility) and the Mixed Oxide Fuel Fabrication Facility is in the preliminary design stage. The WSB is expected to treat a transuranic (TRU) waste stream composed primarily of americium and two low-level waste (LLW) streams. The acidic wastes will be concentrated in the WSB evaporator and neutralized in a cement head tank prior to solidification. A series of TRU mixes were prepared to produce waste forms exhibiting a range of processing and cured properties. The LLW mixes were prepared using the premix from the preferred TRU waste form. All of the waste forms tested passed the Toxicity Characteristic Leaching Procedure. After processing in the WSB, current plans are to dispose of the solidified TRU waste at the Waste Isolation Pilot Plant in New Mexico and the solidified LLW waste at an approved low-level waste disposal facility.

COZZI, ALEX

2004-02-18T23:59:59.000Z

383

A critique of the Department of Energy's environmental programme for the proposed high-level nuclear waste repository at Yucca Mountain, Nevada, USA  

Science Journals Connector (OSTI)

Legislative standards, regulatory provisions, and US Department of Energy (DOE) policies have defined the environmental programme to study the planned high-level nuclear waste repository site at Yucca Mountain, Nevada. The DOE has limited environmental data collection for purposes of monitoring and mitigating impacts of proposed characterisation activities and repository development. Most of the plans are procedural and do not provide for the gathering of baseline data prior to site characterisation activities. The absence of baseline data prior to site characterisation activities reduces the chances of developing a comprehensive and integrated environmental programme to detect and remedy disturbed areas critical to the performance of the repository as well as to assess repository performance properly. The lack of long-term ecological studies means also that interactions between the geohydrologic setting of the repository and the biosphere will not be understood fully.

J. Lemons; D. J. Schaeffer; E. Saboski

1992-01-01T23:59:59.000Z

384

French Geological Repository Project for High Level and Long-Lived Waste: Scientific Programme  

SciTech Connect (OSTI)

The feasibility study presented in the Dossier 2005 Argile set out to evaluate the conditions for building, operating and managing a reversible disposal facility. The research was directed at demonstrating a potential for confining long-lived radioactive waste in a deep clay formation by establishing the feasibility of the disposal principle. Results have been enough convincing and a Planning Act was passed on 28 June, 2006. Decision in principle has been taken to dispose of intermediate and high level long-lived radioactive waste in a geological repository. An application file for a license to construct a disposal facility is requested by end of 2014 and its commissioning is planned for 2025. Based on previous results as well as on recommendations made by various Dossier 2005 evaluators, a new scientific programme for 2006-2015 has been defined. It gives details of what will be covered over the 2006-2015 period. Particular emphasis is placed on consolidating scientific data, increasing understanding of certain mechanisms and using a scientific and technical integration approach. It aims at integrating scientific developments and engineering advances. The scientific work envisaged beyond 2006 has the benefit of a unique context, which is direct access to the geological medium over long timescales. It naturally extends the research carried out to date, and incorporates additional investigations of the geological medium, and the preparation of demonstration work especially through full-scale tests. Results will aim at improving the representation of repository evolutions over time, extract the relevant parameters for monitoring during the reversibility phases, reduce the parametric uncertainties and enhance the robustness of models for performance calculations and safety analyses. Structure and main orientation of the ongoing scientific programme are presented. (author)

Landais, P.; Lebon, P.; Ouzounian, G. [Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA), 92 - Chatenay Malabry (France)

2008-07-01T23:59:59.000Z

385

PROBABILITY BASED CORROSION CONTROL FOR HIGH LEVEL WASTE TANKS: INTERIM REPORT  

SciTech Connect (OSTI)

Controls on the solution chemistry (minimum nitrite and hydroxide concentrations) are in place to prevent the initiation and propagation of pitting and stress corrosion cracking in high level waste (HLW) tanks. These controls are based upon a series of experiments performed on carbon steel coupons in simulated waste solutions. An experimental program was undertaken to investigate reducing the minimum molar nitrite concentration required to confidently inhibit pitting. A statistical basis to quantify the probability of pitting for the tank wall, when exposed to various dilute solutions, is being developed. Electrochemical and coupon testing are being performed within the framework of the statistical test matrix to determine the minimum necessary inhibitor concentrations and develop a quantitative model to predict pitting propensity. A subset of the original statistical test matrix was used to develop an applied understanding of the corrosion response of the carbon steel in the various environments. The interim results suggest that there exists some critical nitrite concentration that sufficiently inhibits against localized corrosion mechanisms due to nitrates/chlorides/sulfates, beyond which further nitrite additions are unnecessary. The combination of visual observation and the cyclic potentiodynamic polarization scans indicate the potential for significant inhibitor reductions without consequence specifically at nitrate concentrations near 1 M. The complete data sets will be used to determine the statistical basis to confidently inhibit against pitting using nitrite inhibition with the current pH controls. Once complete, a revised chemistry control program will be devised based upon the probability of pitting specifically for dilute solutions which will allow for tank specific chemistry control implementation.

Hoffman, E; Karthik Subramanian, K

2008-04-23T23:59:59.000Z

386

RESORCINOL-FORMALDEHYDE ION EXCHANGE RESIN CHEMISTRY FOR HIGH LEVEL WASTE TREATMENT  

SciTech Connect (OSTI)

A principal goal at the Savannah River Site is to safely dispose of the large volume of liquid nuclear waste held in many storage tanks. In-tank ion exchange technology is being considered for cesium removal using a polymer resin made of resorcinol formaldehyde that has been engineered into microspheres. The waste under study is generally lower in potassium and organic components than Hanford waste; therefore, the resin performance was evaluated with actual dissolved salt waste. The ion exchange performance and resin chemistry results are discussed.

Nash, C.; Duignan, M.

2010-01-14T23:59:59.000Z

387

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4  

SciTech Connect (OSTI)

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

Not Available

1994-04-01T23:59:59.000Z

388

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

Broader source: Energy.gov [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.

389

Development of a new generation of waste form for entrapment and immobilization of highly volatile and soluble radionuclides.  

SciTech Connect (OSTI)

The United States is now re-assessing its nuclear waste disposal policy and re-evaluating the option of moving away from the current once-through open fuel cycle to a closed fuel cycle. In a closed fuel cycle, used fuels will be reprocessed and useful components such as uranium or transuranics will be recovered for reuse. During this process, a variety of waste streams will be generated. Immobilizing these waste streams into appropriate waste forms for either interim storage or long-term disposal is technically challenging. Highly volatile or soluble radionuclides such as iodine ({sup 129}I) and technetium ({sup 99}Tc) are particularly problematic, because both have long half-lives and can exist as gaseous or anionic species that are highly soluble and poorly sorbed by natural materials. Under the support of Sandia National Laboratories (SNL) Laboratory-Directed Research & Development (LDRD), we have developed a suite of inorganic nanocomposite materials (SNL-NCP) that can effectively entrap various radionuclides, especially for {sup 129}I and {sup 99}Tc. In particular, these materials have high sorption capabilities for iodine gas. After the sorption of radionuclides, these materials can be directly converted into nanostructured waste forms. This new generation of waste forms incorporates radionuclides as nano-scale inclusions in a host matrix and thus effectively relaxes the constraint of crystal structure on waste loadings. Therefore, the new waste forms have an unprecedented flexibility to accommodate a wide range of radionuclides with high waste loadings and low leaching rates. Specifically, we have developed a general route for synthesizing nanoporous metal oxides from inexpensive inorganic precursors. More than 300 materials have been synthesized and characterized with x-ray diffraction (XRD), BET surface area measurements, and transmission electron microscope (TEM). The sorption capabilities of the synthesized materials have been quantified by using stable isotopes I and Re as analogs to {sup 129}I and {sup 99}Tc. The results have confirmed our original finding that nanoporous Al oxide and its derivatives have high I sorption capabilities due to the combined effects of surface chemistry and nanopore confinement. We have developed a suite of techniques for the fixation of radionuclides in metal oxide nanopores. The key to this fixation is to chemically convert a target radionuclide into a less volatile or soluble form. We have developed a technique to convert a radionuclide-loaded nanoporous material into a durable glass-ceramic waste form through calcination. We have shown that mixing a radionuclide-loaded getter material with a Na-silicate solution can effectively seal the nanopores in the material, thus enhancing radionuclide retention during waste form formation. Our leaching tests have demonstrated the existence of an optimal vitrification temperature for the enhancement of waste form durability. Our work also indicates that silver may not be needed for I immobilization and encapsulation.

Rodriguez, Mark Andrew; Bencoe, Denise Nora; Brinker, C. Jeffrey; Murphy, Andrew Wilson; Holt, Kathleen Caroline; Turnham, Rigney; Kruichak, Jessica Nicole; Tellez, Hernesto; Miller, Andy; Xiong, Yongliang; Pohl, Phillip Isabio; Ockwig, Nathan W.; Wang, Yifeng; Gao, Huizhen

2010-09-01T23:59:59.000Z

390

E-Print Network 3.0 - activity-long life waste Sample Search...  

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

ASME Proceedings of NAWTEC16 Summary: to as an "environmental life-cycle assessment," or LCA." In the field of solid waste management, it has long been... . The conduct of life...

391

Independent Oversight Activity Report, Hanford Waste Treatment and Immobilization Plant – November 2013  

Broader source: Energy.gov [DOE]

Catholic University of America Vitreous State Laboratory Tour and Discussion of Experiments Conducted in Support of Hanford Site Waste Treatment and Immobilization Plant Select Systems Design [HIAR-VSL-2013-11-18

392

Independent Activity Report, Office of River Protection Waste Treatment Plant and Tank Farms- February 2013  

Broader source: Energy.gov [DOE]

Site Familiarization and Introduction of New Office of Safety and Emergency Management Evaluations Site Lead for the Office of River Protection Waste Treatment Plant and Tank Farms [HIAR-HANFORD-2013-02-25

393

E-Print Network 3.0 - activity waste vitrification Sample Search...  

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

Conference May 1-3, 2006, Tampa, Florida USA Summary: -98. Izumikawa C., 1996, "Metal recovery from fly ash generated from vitrification process for MSW ash," Waste... 14th...

394

Electrochemical determination of the corrosion behavior of candidate alloys proposed for containment of high level nuclear waste in tuff  

SciTech Connect (OSTI)

Long-term geological disposal of nuclear waste requires corrosion-resistant canister materials for encapsulation. Several austenitic stainless steels are under consideration for such purposes for the disposal of high-level waste at the candidate repository site located at Yucca Mountain, Nevada. With regard to corrosion considerations, a worst case scenario at this prospective repository location would result from the intrusion of vadose water. This preliminary study focuses on the electrochemical and corrosion behavior of the candidate canister materials under worst-case repository environments. Electrochemical parameters related to localized attack (e.g., pitting potentials) and the electrochemical corrosion rates have been examined. 15 references, 15 figures, 4 tables.

Glass, R.S.; Overturf, G.E.; Garrison, R.E.; McCright, R.D.

1984-06-18T23:59:59.000Z

395

Active control system for high speed windmills  

DOE Patents [OSTI]

A pump stroke is matched to the operating speed of a high speed windmill. The windmill drives a hydraulic pump for a control. Changes in speed of a wind driven shaft open supply and exhaust valves to opposite ends of a hydraulic actuator to lengthen and shorten an oscillating arm thereby lengthening and shortening the stroke of an output pump. Diminishing wind to a stall speed causes the valves to operate the hydraulic cylinder to shorten the oscillating arm to zero. A pressure accumulator in the hydraulic system provides the force necessary to supply the hydraulic fluid under pressure to drive the actuator into and out of the zero position in response to the windmill shaft speed approaching and exceeding windmill stall speed. 4 figs.

Avery, D.E.

1988-01-12T23:59:59.000Z

396

Management activities for retrieved and newly generated transuranic waste, Savannah River Plant  

SciTech Connect (OSTI)

The purpose of this Environmental Assessment (EA) is to assess the potential environmental impacts of the retrieval and processing of retrieved and newly generated transuranic (TRU) radioactive waste at the Savannah River Plant (SRP), including the transportation of the processes TRU waste to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. A new TRU Waste Facility (TWF) will be constructed at SRP to retrieve and process the SRP TRU waste in interim storage to meet WIPP criteria. This EA has been prepared in compliance with the National Environmental Policy Act (NEPA) of 1969, as amended, and the requirements of the Council of Environmental Quality Regulations for implementing NEPA (40 CFR Parts 1500--1508). The National Environmental Policy Act (NEPA) requires the assessment of environmental consequences of all major federal actions that may affect the quality of the human environment. This document describes the environmental impact of constructing and operating the TWF facility for processing and shipment of the TRU waste to WIPP and considers alternatives to the proposed action. 40 refs., 12 figs., 12 tabs.

Not Available

1988-08-01T23:59:59.000Z

397

THERMAL DESTRUCTION OF HIGHLY CHLORINATED MIXED WASTES WITHOUT GENERATING CORROSIVE OFF-GASES USING MOLTEN SALT OXIDATION (1,2)  

SciTech Connect (OSTI)

A pilot-scale MSO (Molten Salt Oxidation) system was used to process 45-gallons of a halogenated mixed waste that is difficult to treat with other thermal systems. The mixed waste was a halogenated solvent that consisted mostly of methylchloroform. The 80 weight percent of waste consisting of highly corrosive chlorine was captured in the first process vessel as sodium chloride. The sodium chloride leached chrome from that process vessel and the solidified salt exhibited the toxicity characteristic for chrome as measured by TCLP (Toxicity Characteristic Leaching Procedure) testing. The operating ranges for parameters such as salt bed temperature, off-gas temperature, and feed rate that enable sustained operation were identified. At feed rates below the sustainable limit, both processing capacity and maintenance requirements increased with feed rate. Design and operational modifications to increase the sustainable feed rate limit and reduce maintenance requirements reduced both salt carryover and volumetric gas flows.

Smith, W.; Feizollahi, F.

2002-02-25T23:59:59.000Z

398

The Transmutation of Nuclear Waste in the Two-Zone Subcritical System Driven by High- Intensity Neutron Generator - 12098  

SciTech Connect (OSTI)

The main problems of transmutation of high-level radioactive waste (minor actinides and long-lived fission products) are considered in our work. The range of radioactive waste of nuclear power is analyzed. The conditions under which the transmutation of radioactive waste will be most effective are analyzed too. The modeling results of a transmutation of the main radioactive isotopes are presented and discussed. The transmutation of minor actinides and long-lived fission products are modeled in our work (minor actinides - Np-237, Am-241, Am-242, Am-243, Cm-244, Cm-245; long-lived fission products - I-129, Tc-99). The two-zone subcritical system is calculated with help of different neutron-physical codes (MCNP, Scale, Montebarn, Origen). The ENDF/B-VI nuclear data library used in above calculations. Thus, radioactive wastes can be divided into two main groups that need to be transmuted. The minor actinides form the first group and the long-lived fission products form the second one. For the purpose of effective transmutation these isotopes must be extracted from the spent nuclear fuel with the help of either PUREX technology or pyrometallurgical technology. The two-zone reactor system with fast and thermal regions is more effective for nuclear waste transmutation than the one-zone reactor. Modeling results show that nearly all radioactive wastes can be transmuted in the two-zone subcritical system driven by a high-intensity neutron generator with the external neutron source strength of 1.10{sup 13} n/sec. Obviously, transmutation rate will increase with a rise of the external neutron source strength. From the results above we can also see that the initial loading of radioactive isotopes into the reactor system should exceed by mass those isotopes that are finally produced. (authors)

Babenko, V.O. [Bogolyubov Institute for Theoretical Physics, Metrolohichna str. 14-b, Kiev, 03680 (Ukraine); Gulik, V.I.; Pavlovych, V.M. [Institute for Nuclear Research, pr. Nauky 47, Kyiv, 03680 (Ukraine)

2012-07-01T23:59:59.000Z

399

A data base and a standard material for use in acceptance testing of low-activity waste products  

SciTech Connect (OSTI)

The authors have conducted replicate dissolution tests following the product consistency test (PCT) procedure to measure the mean and standard deviation of the solution concentrations of B, Na, and Si at various combinations of temperature, duration, and glass/water mass ratio. Tests were conducted with a glass formulated to be compositionally similar to low-activity waste products anticipated for Hanford to evaluate the adequacy of test methods that have been designated in privatization contracts for use in product acceptance. An important finding from this set of tests is that the solution concentrations generated in tests at 20 C will likely be too low to measure the dissolution rates of waste products reliably. Based on these results, the authors recommend that the acceptance test be conducted at 40 C. Tests at 40 C generated higher solution concentrations, were more easily conducted, and the measured rates were easily related to those at 20 C. Replicate measurements of other glass properties were made to evaluate the possible use of LRM-1 as a standard material. These include its composition, homogeneity, density, compressive strength, the Na leachability index with the ANSI/ANS 16.1 leach test, and if the glass is characteristically hazardous with the toxicity characteristic leach procedure. The values of these properties were within the acceptable limits identified for Hanford low-activity waste products. The reproducibility of replicate tests and analyses indicates that the glass would be a suitable standard material.

Wolf, S.F.; Ebert, W.L.; Luo, J.S.; Strachan, D.M.

1998-04-01T23:59:59.000Z

400

Nuclear Waste Disposal: Amounts of Waste  

Science Journals Connector (OSTI)

The term nuclear waste...embraces all residues from the use of radioactive materials, including uses in medicine and industry. The most highly radioactive of these are the spent fuel or reprocessed wastes from co...

2005-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "high activity waste" 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.


401

The Northwest Geysers High-Temperature Reservoir- Evidence For Active  

Open Energy Info (EERE)

Geysers High-Temperature Reservoir- Evidence For Active Geysers High-Temperature Reservoir- Evidence For Active Magmatic Degassing And Implications For The Origin Of The Geysers Geothermal Field Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: The Northwest Geysers High-Temperature Reservoir- Evidence For Active Magmatic Degassing And Implications For The Origin Of The Geysers Geothermal Field Details Activities (2) Areas (1) Regions (0) Abstract: Noble gas isotope abundances in steam from the Coldwater Creek field of the Northwest Geysers, California, show mixing between a nearly pure mid-ocean ridge (MOR) type magmatic gas with high 3He/4He and low radiogenic 40*Ar (R/Ra > 8.3 and 40*Ar/4He < 0.07), and a magmatic gas diluted with crustal gas (R/Ra 0.25). The

402

High-Activity Radioactive Materials Removed From Mexico | National Nuclear  

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

High-Activity Radioactive Materials Removed From Mexico | National Nuclear High-Activity Radioactive Materials Removed From Mexico | 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 > Media Room > Press Releases > High-Activity Radioactive Materials Removed From Mexico Press Release High-Activity Radioactive Materials Removed From Mexico Nov 15, 2013 WASHINGTON, D.C. - The National Nuclear Security Administration (NNSA)

403

High-Activity Radioactive Materials Removed From Mexico | National Nuclear  

National Nuclear Security Administration (NNSA)

High-Activity Radioactive Materials Removed From Mexico | National Nuclear High-Activity Radioactive Materials Removed From Mexico | 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 > Media Room > Press Releases > High-Activity Radioactive Materials Removed From Mexico Press Release High-Activity Radioactive Materials Removed From Mexico Nov 15, 2013 WASHINGTON, D.C. - The National Nuclear Security Administration (NNSA)

404

High Level Overview of DOE Biomass Logistics II Project Activities  

Broader source: Energy.gov [DOE]

Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers High Level Overview of DOE Biomass Logistics II Project Activities Kevin Comer, Associate Principal, Antares Group Inc.

405

NNSA Removes High-Activity Radioactive Materials from Boston | National  

National Nuclear Security Administration (NNSA)

Removes High-Activity Radioactive Materials from Boston | National Removes High-Activity Radioactive Materials from Boston | 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 > Media Room > Press Releases > NNSA Removes High-Activity Radioactive Materials from Boston Press Release NNSA Removes High-Activity Radioactive Materials from Boston Nov 22, 2013

406

Nuclear waste solids  

Science Journals Connector (OSTI)

Glass and polycrystalline materials for high-level radioactive waste immobilization are discussed. Borosilicate glass has been selected as the waste form for defence high-level radwaste in the US. Since releas...

L. L. Hench; D. E. Clark; A. B. Harker

1986-05-01T23:59:59.000Z

407

Measurement of radioactive contaminated wastes  

SciTech Connect (OSTI)

At Los Alamos, a comprehensive program is underway for the development of sensitive, practical, nondestructive assay techniques for the quantification of low-level transuranics in bulk solid wastes. The program encompasses a broad range of techniques, including sophisticated active and passive gamma-ray spectroscopy, passive neutron detection systems, pulsed portable neutron generator interrogation systems, and electron accelerator-based techniques. The techniques can be used with either low-level or high-level beta-gamma wastes in either low-density or high-density matrices. The techniques are quite sensitive (< 10 nCi/g detection) and, in many cases, isotopic specific. Waste packages range in size from small cardboard boxes to large metal or wooden crates. Considerable effort is being expended on waste matrix identification to improve assay accuracy.

Caldwell, J.T.; Close, D.A.; Crane, T.W.

1983-01-01T23:59:59.000Z

408

Combustion of high-sulfur coal and anthracite wastes in a rotary kiln combustor with an advanced internal air distributor  

SciTech Connect (OSTI)

Fluid bed combustors have received extensive testing with both high-sulfur coal and anthracite wastes. Rotary kilns are effective and popular devices for waste combustion. The Angelo Rotary Furnace{trademark} has been developed to improve the operation of rotary pyrolyzer/combustor systems through enhanced air distribution, which in this process is defined as staged, swirled combustion air injection. Fourteen of these new furnaces have been installed worldwide. Two units in Thailand, designed for rice hull feed with occasional lignite feed, have been recently started up. An older unit in Pennsylvania is being upgraded with a new, more advanced air distribution system for a series of tests this fall in which inexpensive high-sulfur coal and anthracite wastes will be fired with limestone. The purposes of these tests are to determine the burning characteristics of these two fuels in this system, to discover the Ca/S ratios necessary for operation of a rotary kiln combusting these fuels, and to observe the gas-borne emissions from the furnace. An extensive preliminary design study will be performed on a commercial installation for combustion of anthracite wastes. 14 refs., 5 figs., 1 tab.

Cobb, J.T. Jr. (Pittsburgh Univ., PA (USA)); Ahn, Y.K. (Gilbert/Commonwealth, Inc., Reading, PA (USA)); Angelo, J.F. (Universal Energy International, Inc., Little Rock, AR (USA))

1990-01-01T23:59:59.000Z

409

Radiological assessment of the consequences of the disposal of high-level radioactive waste in subseabed sediments  

SciTech Connect (OSTI)

The radiological assessment of the seabed option consists in estimating the detriment to man and to the environment that could result from the disposal of high-level waste (HLW) within the seabed sediments in deep oceans. The assessment is made for the high-level waste (vitrified glass) produced by the reprocessing of 10/sup 5/ tons of heavy metal from spent fuel, which represents the amount of waste generated by 3333 reactor-yr of 900-MW(electric) reactors, i.e., 3000 GW(electric) x yr. The disposal option considered is to use 14,667 steel penetrators, each of them containing five canisters of HLW glass (0.15 m/sup 3/ each). These penetrators would reach a depth of 50 m in the sediments and would be placed at an average distance of 180 m from each other, requiring a disposal area on the order of 22 x 22 km. Two such potential disposal areas in the Atlantic Ocean were studied, Great Meteor East (GME) and South Nares Abyssal Plains (SNAP). A special ship design is proposed to minimize transportation accidents. Approximately 100 shipments would be necessary to dispose of the proposed amount of waste. The results of this radiological assessment seem to show that the disposal of HLW in subseabed sediments is radiologically a very acceptable option.

de Marsily, G.; Behrendt, V.; Ensminger, D.A.; Flebus, C.; Hutchinson, B.L.; Kane, P.; Karpf, A.; Klett, R.D.; Mobbs, S.; Poulin, M.; Stanners, D.A.; Wuschke, D.

1987-01-01T23:59:59.000Z

410

Radiological assessment of the consequences of the disposal of high level radioactive waste in sub-seabed sediments  

SciTech Connect (OSTI)

The radiological assessment of the Seabed option consists of estimating the detriment to man and to the environment that could result from the disposal of high-level nuclear waste within the seabed sediments in the deep oceans. The assessment is made for the high-level waste (vitrified glass) produced by the reprocessing of 10/sup 5/ tons of heavy metal from spent fuel, which represents the amount of waste generated by 3333 reactor-years of 900 MW(e) reactors, i.e., 3000 GW(E).a. The disposal option considered is to use 14,667 steel penetrators, each of them containing five canisters of HLW glass (0,15 m/sup 3/ each). These penetrators would reach a depth of 50 m in the sediments and would be placed at an average distance of 180 m from each other, requiring a disposal area on the order of 22 x 22 km. Two such potential disposal areas in the Atlantic ocean were studied, Great Meteor East (GME) and South Nares Abyssal Plains (SNAP). A special ship design is proposed to minimize transportation accidents. Approximately 100 shipments would be necessary to dispose of the proposed amount of waste. 1 ref.

de Marsily, G.; Behrendt, V.; Ensminger, D.A.; Flebus, C.; Hutchinson, B.L.; Kane, P.; Karpf, A.; Klett, R.D.; Mobbs, S.; Poulin, M.

1987-01-01T23:59:59.000Z

411

US Department of Energy Storage of Spent Fuel and High Level Waste  

SciTech Connect (OSTI)

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

412

Evaluation of DMDOHEMA based supported liquid membrane system for high level waste remediation under simulated conditions  

Science Journals Connector (OSTI)

Abstract N,N?-dimethyl-N,N?-dioctyl-2,(2?-hexyloxyethyl) malonamide (DMDOHEMA) has been proposed as solvent for the partitioning of radiotoxic minor actinides from high-level waste (HLW) solutions. The facilitated transport of 241Am(III), 239Pu(IV), 233U(VI), 237Np(V) across supported liquid membrane (SLM) impregnated with DMDOHEMA solution in n-dodecane was investigated under varying conditions of feed acidity, receiver phase composition, carrier concentration, and membrane thickness. Micro porous PTFE membrane was used as the polymeric support. There was a decrease in the transport of metal ions under the pressurized heavy water reactor simulated HLW (PHWR-SHLW) conditions. The physical stability of the SLM impregnated with the carrier was investigated for ~60 days by performing Am(III) permeation studies. Marginal variation in the transport behavior suggested reasonably good stability of the impregnated carrier in the membrane pores. A simple mathematical model has been developed to simulate experimental data and to explain quantitatively the role of different parameters.

Ajay B. Patil; Pankaj Kandwal; V.S. Shinde; P.N. Pathak; P.K. Mohapatra

2013-01-01T23:59:59.000Z

413

Kinetic model for quartz and spinel dissolution during melting of high-level-waste glass batch  

SciTech Connect (OSTI)

The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-level wastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter.

Pokorny, Richard; Rice, Jarrett A.; Crum, Jarrod V.; Schweiger, Michael J.; Hrma, Pavel R.

2013-07-24T23:59:59.000Z

414

Sedimentation behavior of noble metal particles in simulated high-level waste borosilicate glasses  

SciTech Connect (OSTI)

Solubility of noble metal elements (NME) in the melted borosilicate glass is much smaller than its normal concentration of the high level liquid waste. Thus most of NME show small particles in the melted glass and tend to sediment in the bottom region of the vitrification melter due to their higher density than that of glass. Experiments of the sedimentation of NME particles in the melted glass were carried out under static condition. Three conditions of initial NME concentration (1.1, 3.0, 6.1 wt % with an equivalent for each oxide) in the simulated glass were set and held at 1100 C. degrees up to 2880 hours. The specimen with 1.1 wt % initial NME concentration indicated zone settling, and the settling rate of the interface is constant: 2.4 mm/h. This sedimentation behavior is the type of rapid settling. Following the rapid settling, the settling rate goes gradually slower; this is the type of compressive settling. The specimens with 3.0 wt % and 6.1 wt % initial NME concentration showed compression settling from the beginning. From the settling curve of the interface, the maximum concentration of NME in sediment was estimated to be around 23- 26 wt %. Growth of NME particles was observed by holding at 1100 C. degrees for up to 2880 hours. The viscosity becomes higher as NME concentration increases and the dependence on shear rate becomes simultaneously stronger. The effect of the particle growth to viscosity appears to be not significant.

Nakajima, M.; Ohyama, K.; Morikawa, Y.; Miyauchi, A.; Yamashita, T. [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1109 (Japan); Komamine, S.; Ochi, E. [Japan Nuclear Fuel Limited, Bussan-Bldg. Bekkan, 1-1-5 Nishi-Shinbashi Minato-ku, Tokyo 105-0003 (Japan)

2013-07-01T23:59:59.000Z

415

Oxidative Alkaline leaching of Americium from simulated high-level nuclear waste sludges  

SciTech Connect (OSTI)

Oxidative alkaline leaching has been proposed to pre-treat the high-level nuclear waste sludges to remove some of the problematic (e.g., Cr) and/or non-radioactive (e.g., Na, Al) constituents before vitrification. It is critical to understand the behavior of actinides, americium and plutonium in particular, in o