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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Application of 'Six Sigma{sup TM}' and 'Design of Experiment' for Cementation - Recipe Development for Evaporator Concentrate for NPP Ling AO, Phase II (China) - 12555

Cementation of radioactive waste is a common technology. The waste is mixed with cement and water and forms a stable, solid block. The physical properties like compression strength or low leach ability depends strongly on the cement recipe. Due to the fact that this waste cement mixture has to fulfill special requirements, a recipe development is necessary. The Six Sigma{sup TM}' DMAIC methodology, together with the Design of experiment (DoE) approach, was employed to optimize the process of a recipe development for cementation at the Ling Ao nuclear power plant (NPP) in China. The DMAIC offers a structured, systematical and traceable process to derive test parameters. The DoE test plans and statistical analysis is efficient regarding the amount of test runs and the benefit gain by getting a transfer function. A transfer function enables simulation which is useful to optimize the later process and being responsive to changes. The DoE method was successfully applied for developing a cementation recipe for both evaporator concentrate and resin waste in the plant. The key input parameters were determined, evaluated and the control of these parameters were included into the design. The applied Six Sigma{sup TM} tools can help to organize the thinking during the engineering process. Data are organized and clearly presented. Various variables can be limited to the most important ones. The Six Sigma{sup TM} tools help to make the thinking and decision process trace able. The tools can help to make data driven decisions (e.g. C and E Matrix). But the tools are not the only golden way. Results from scoring tools like the C and E Matrix need close review before using them. The DoE is an effective tool for generating test plans. DoE can be used with a small number of tests runs, but gives a valuable result from an engineering perspective in terms of a transfer function. The DoE prediction results, however, are only valid in the tested area. So a careful selection of input parameter and their limits for setting up a DoE is very important. An extrapolation of results is not recommended because the results are not reliable out of the tested area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Numerical study of radiation effect on the municipal solid waste combustion characteristics inside an incinerator

Highlights: • A 3-D model for the MSW incinerator with preheated air was developed. • Gas radiative properties were obtained from a statistical narrow-band model. • Non-gray body radiation model can provide more accurate simulation results. - Abstract: Due to its advantages of high degree volume reduction, relatively stable residue, and energy reclamation, incineration becomes one of the best choices for Municipal Solid Waste (MSW) disposal. However, detailed measurements of temperature and gas species inside a furnace are difficulty by conventional experimental techniques. Therefore, numerical simulation of MSW incineration in the packed bed and gas flow field was applied. In this work, a three dimensional (3-D) model of incinerator system, including flow, heat transfer, detailed chemical mechanisms, and non-gray gas models, was developed. Radiation from the furnace wall and the flame formed above the bed is of importance for drying and igniting the waste. The preheated air with high temperature is used for the MSW combustion. Under the conditions of high temperature and high pressure, MSW combustion produces a variety of radiating gases. The wavelength-depend radiative properties of flame adopted in non-gray radiation model were obtained from a statistical narrow-band model. The influence of radiative heat transfer on temperature, flow field is researched by adiabatic model (without considering radiation), gray radiation model, and non-gray radiation model. The simulation results show that taking into account the non-gray radiation is essential.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Experimental investigation on heat transfer performance of drain water heat recovery heat exchangers

Water heating accounts for 18 % of residential energy consumption, and a substantial portion of this energy can be recovered through drain water heat recovery systems. Here, this paper investigates the heat transfer performance of drain water heat recovery heat exchangers. Two heat exchanger configurations were studied: vertical (gravity film exchanger) and horizontal. The heat recovery performance of the exchangers was quantified using sensible effectiveness and characteristic curves following the ε-NTU approach. Depending on the flow conditions, heat exchangers with a vertical configuration offered higher heat transfer effectiveness compared with the horizontal configuration under similar operating conditions. Furthermore, the results showed that properly designed and sized heat exchangers can effectively recover considerable amounts of waste heat from drains, depending on flow rates. Additionally, in the vertical heat exchanger configuration, low flow rates result in partially wet conditions, resulting in lower heat transfer effectiveness. Therefore, two distributor configurations are proposed to improve the heat transfer performance of vertically configured heat exchangers at low flow rates. The proposed distributor configurations and data presented in this paper will be useful in the design and development of drain water heat recovery systems in building applications.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2023

Analysis of the Sludge Batch 9 (Macrobatch 11) DWPF Pour Stream Glass Sample

A pour stream (PS) sample taken near the end of processing of Sludge Batch 9 (SB9, Macrobatch 11) with the Nitric-Formic Acid Flowsheet was analyzed by the Savannah River National Laboratory to comply with the Defense Waste Processing Facility (DWPF) Glass Product Control Program. The SB9 PS sample was taken on 5/28/2022 during Melter Feed Batch 815 associated with glass canister S04741, and was transferred to SRNL in Primary Container PC0136.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

REPORTABLE RADIONUCLIDES IN DWPF SLUDGE BATCH 7A (MACROBATCH 8)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that the waste producer 'shall report the curie inventory of radionuclides that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115.' As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type all radionuclides that have half-lives longer than 10 years and contribute greater than 0.01 percent of the total curie inventory from the time of production through the 1100 year period from 2015 through 3115. The initial list of radionuclides to be reported is based on the design-basis glass identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report. However, it is required that the list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the 'greater than 0.01% of the curie inventory' criterion. Specification 1.6 of the WAPS, International Atomic Energy Agency Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, and U-238; and Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete list of reportable radionuclides must also include these sets of U and Pu isotopes - and the U and Pu isotopic mass distributions must be identified. The DWPF receives HLW sludge slurry from Savannah River Site (SRS) Tank 40. For Sludge Batch 7a (SB7a), the waste in Tank 40 contained a blend of the heel from Sludge Batch 6 (SB6) and the Sludge Batch 7 (SB7) material transferred to Tank 40 from Tank 51. This sludge blend is also referred to as Macrobatch 8. Laboratory analyses of a Tank 40 sludge sample were performed to quantify the concentrations of pertinent radionuclides in the SB7a waste. Subsequently, radiological decay and in-growth were calculated over the time period from 2015 to 3115. This provided a basis for characterizing the radionuclide content of SB7a over time and for identifying the 'reportable radionuclides.' Details of the characterization methodology and the analytical results are the focus of this report. This work was performed at the request of the Waste Solidification Engineering Department of Savannah River Remediation, initiated via Technical Task Request (TTR) HLW-DWPF-TTR-2010-0031. A minor revision in the reporting requirements was requested via a subsequent email communication. The work was conducted in accordance with the protocols identified in Task Technical and Quality Assurance Plan SRNL-RP-2010-01218 and Analytical Study Plan SRNL-RP-2010-01219. All of the raw data related to this scope have been recorded in laboratory notebook SRNL-NB-2011-00061. The overall goal of this task was to characterize the radionuclide content of the SB7a waste sufficiently to meet the WAPS and DWPF reporting requirements. The specific objectives were: (1) Quantify the current concentrations of all radionuclides impacting (or potentially-impacting) the total curie content between calendar years 2011 and 3115. Also quantify the current concentrations of other radionuclides specifically requested in the TTR or required by the WAPS. (2) Calculate future concentrations of decayed and in-grown radionuclides impacting the total curie content between calendar years 2015 and 3115; (3) Identify as 'reportable' all radionuclides contributing {ge} 0.01% of the total curie content from 2015 to 3115 and having half-lives {ge} 10 years.

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DOE OSTI2020

Performance Prediction of Big Data Transfer Through Experimental Analysis and Machine Learning

Big data transfer in next-generation scientific ap- plications is now commonly carried out over connections with guaranteed bandwidth provisioned in High-performance Net- works (HPNs) through advance bandwidth reservation. To use HPN resources efficiently, provisioning agents need to carefully schedule data transfer requests and allocate appropriate band- widths. Such reserved bandwidths, if not fully utilized by the requesting user, could be simply wasted or cause extra overhead and complexity in management due to exclusive access. This calls for the capability of performance prediction to reserve bandwidth resources that match actual needs. Towards this goal, we employ machine learning algorithms to predict big data transfer perfor- mance based on extensive performance measurements, which are collected over a span of several years from a large number of data transfer tests using different protocols and toolkits between various end sites on several real-life physical or emulated HPN testbeds. We first identify a comprehensive list of attributes involved in a typical big data transfer process, including end host system configurations, network connection properties, and control parameters of data transfer methods. We then conduct an in-depth exploratory analysis of their impacts on application- level throughput, which provides insights into big data transfer performance and motivates the use of machine learning. We also investigate the applicability of machine learning algorithms and derive their general performance bounds for performance prediction of big data transfer in HPNs. Experimental results show that, with appropriate data preprocessing, the proposed machine learning-based approach achieves 95% or higher pre- diction accuracy in up to 90% of the cases with very noisy real-life performance measurements.

Yun, Daqing↗
DOE OSTI2019

The Applications of Intumescent Technologies in Support of D and D Activities across the DOE Complex - 19599

The development of intumescent technologies to be used as both an excellent means of fire protection and resiliency and as a radioactive fixative, would most definitely benefit the safety of the general public and also the environment. The remediation and safe disposal of radioactive waste is of great important to both national security and the public health, as we carry the responsibility to not develop new ways of mitigating these hazards but also to lay the groundwork for future generations. The overall goal of this research is to enhance the fire resiliency and protection of fixatives and nuclear facilities in support of D and D activities. Specific objectives include: 1) identify and evaluate commercial-off-the-shelf (COTS) intumescent technologies, 2) conduct baseline fire testing of intumescent technologies, 3) develop a set of international standards to meet operational and safety requirements for D and D Technologies. In order to evaluate the above proposed research objectives, the following hypotheses have been postulated. - Select, test, and evaluate the suitability of certain 'near fit' ASTM flammability standards. - Baseline and down-select a set of existing COTS intumescent technologies (coatings and foams) to support D and D problem sets. - Determine feasibility of application methods for use inside a radioactive glovebox and hot cells for D and D purposes. - Develop international testing protocols for adapting the use of intumescent technologies to address needs of the D and D process. The proposed research will primarily take place at FIU-ARC. ARC and SRNL have conducted extensive coordination with the Savannah River Site's (SRS) 235-F PuFF facility personnel to facilitate planning in order to expedite the research and development (R and D) related to this effort. An initial set of documents associated with the specific facility hot cells, including schematics, assessment of the manipulator arms, anticipated particulate sizes of the Pu-238 contamination, an approved tooling list, etc., were formally reviewed and approved for release by the site to ARC. Utilizing these as a foundation, pre-planning for further adaptation of the intumescent technologies and the application during a full-scale cold demo at a hot cell testbed at ARC. One of the primary objectives is the testing and evaluation of the resistance of radiological shielding foams to thermal stressors. This objective was accomplished by conducting a flammability test and incremental temperature test. The goal of the flammability test is to conduct analytical tests to provide a classification for the foams as outlined below in accordance with IEC 60695-11-10. This test method is uniformly accepted throughout the industry as being critical indicators that quantifiably establish the flammability characteristics of plastics. It should be noted that polyurethane is not classified as a plastic, however, they do share similar properties deeming this a near fit test. IEC 60695-11-10:2013 specifies small-scale laboratory test procedures intended to compare the burning behavior of different materials when vertically or horizontally oriented test bar specimens are exposed to a small flame ignition source. Two test methods are described. Method A is a horizontal burning test and is intended to determine the linear burning rate of materials under specific test conditions. Method B is a vertical burning test and is intended to determine whether materials self-extinguish under specific test conditions. The purpose of the incremental temperature testing is to evaluate mass loss and dimensional changes that occurs to the radiological shielding foams when exposed to high temperatures in a muffle furnace. High temperatures often cause materials to undergo physical changes such as swelling or expanding. This is caused by the increase in kinetic energy and atomic and molecular movement.

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NASA NTRS1991

Nonazeotropic Heat Pump

Heat pump collects heat from water circulating in heat-rejection loop, raises temperature of collected heat, and transfers collected heat to water in separate pipe. Includes sealed motor/compressor with cooling coils, evaporator, and condenser, all mounted in outer housing. Gradients of temperature in evaporator and condenser increase heat-transfer efficiency of vapor-compression cycle. Intended to recover relatively-low-temperature waste heat and use it to make hot water.

Ealker, David H.↗
DOE OSTI2023

Novel waste-derived non-noble metal catalysts to accelerate acidic and alkaline hydrogen evolution reaction

Carbon substrate support is a promising approach to facilitate catalytic activity of non-noble metal catalysts in hydrogen evolution reactions (HER). Here, in this paper, we report a waste-derived electrocatalyst P-CoNiO 2 @NC synthesized under mild condition. It exhibits outstanding HER performance in both acid (η 10 = 64.6 mV) and alkaline (η 10 = 96.4 mV) condition with good stability. Especially, it begins to outperform commercial Pt/C when the current density is over 74 mA/cm 2 in acid electrolyte and 29 mA/cm 2 in alkaline electrolyte. Moreover, the waste-derived carbon substrate could effectively promote HER activity of other dual 3d TMOs such as Fe and Co, Fe and Ni. The improved HER activity originates from electron transfer, electronic structure adjustment and optimization of intermediates adsorption according to X -ray photoelectron spectroscopy and density functional theory calculations. This work provides a new insight for converting wastes into value-added products and a sustainable and friendly strategy to facilitate hydrogen energy production.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration

The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumed constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5 Degree-Sign C and the peaks of temperature occurred with less than 8 h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5 Degree-Sign C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Application of evolved gas analysis to cold-cap reactions of melter feeds for nuclear waste vitrification

The melter feed (a mixture of nuclear waste and glass forming and modifying additives) experiences multiple gas-evolving reactions in an electrical glass-melting furnace during the vitrification of nuclear wastes. Foams from the residual gases can significantly alter the melting rate through mass and heat transfers. Here, we employed the thermogravimetry-gas chromatography-mass spectrometry (TGA-GC-MS) combination to perform quantitative evolved gas analysis (EGA) and developed a simple calibration model which correlates the overall mass loss rate with the evolution rates for individual gases. The model parameters are obtained from the least squares analysis, assuming that the gas-evolving reactions are independent. Finally, the EGA adds the ‘chemical identity’ to the reactions indicated by the ‘phenomenological’ kinetic model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Corrosion Evaluation for Tank 241-AP-106 Re-purposing

Tank AP-106 will be an integral part of the Direct Feed Low Activity Waste (DF-LAW) activity. This tank is an underground, carbon steel, double shell tank (DST). Before it can be utilized, the Cs-137 concentration levels in the waste must be reduced significantly. The AP-106 re-purposing process will manage normal waste tank additions procedures such that waste acceptance criteria objectives for Cs-137 removal are achieved. As this process was being developed, concerns regarding tank integrity during and after the process to Tank AP-106 and associated tanks (i.e., AP-102, AN-101, and AP-108) were addressed. The degradation mechanisms identified include: general and pitting corrosion, caustic stress corrosion cracking, galvanic corrosion, and microbiologically induced corrosion (MIC). Recommendations were provided to minimize the exposure of the tank to corrosive conditions (see Section 6.0). These recommendations included chemistry adjustments at defined steps in the process that will maintain the waste in compliance with the corrosion control program. Additionally, sampling and inspection (see Sections 6.0 and 7.0, respectively) recommendations that will assess the final condition of the tank environment and the tank were presented. The greatest risk for this process involves the handling of large quantities of Columbia River water (i.e., greater than 450,000 gallons) within the AP-106, associated tanks and supporting infrastructure (i.e., transfer line piping). As long as this water is blended into the supernate in a timely fashion, and the waste is adjusted appropriately with corrosion inhibitors, the final condition is expected to be within compliance. Recommendations were also made to address process interruptions that could result in exposure of the tank walls to aggressive conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

A heat and mass transfer model for evaluation of damaged cesium chloride radiological sources

Radiological sources are vital to many applications, and typically contain byproducts from waste streams which decay over time to form complex mixtures of elements. Daughter products resulting from these decay processes can introduce complicating factors to the integrity and safety of the vessel that contains the radiological material. Sources containing cesium 137 ( 137 Cs) are of particular interest, because the decay of this isotope produces barium metal which reacts readily and exothermically with oxygen. This work employs physics-based numerical tools to examine the thermal response of a radiological source containing a mixture of cesium and barium in the event that vessel walls are damaged and atmospheric gases contact source material. A parametric study was conducted to determine the sensitivity of the response to various factors, including vessel geometry, source material age, the degree of vessel damage, and other parameters. Here, it was found that the peak temperatures that occurred within the source material strongly depend on these parameters, particularly vessel geometry and age, which determine whether a breach would be a relatively minor accident or a catastrophic incident. Finally, the model’s sensitivity to uncertain thermophysical properties is discussed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Reduced-Order CFD Modeling to Support Waste Loading Optimization in Hanford WTP Vitrification

The U.S. DOE Hanford Site stores over 56 million gallons of radioactive liquid tank waste that must be treated and immobilized for long-term disposal The Waste Treatment and Immobilization Plant (WTP) will vitrify this waste by feeding it into Joule-heated melters, where it is incorporated into a stable borosilicate glass Computational fluid dynamics (CFD) simulations of glass melters can provide insight into the maximum achievable waste loading under varying melter operating conditions Fully resolved VOF multiphase simulations were used as the reference model to capture bubble-driven convection in the melter, including bubble formation, rise behavior, and induced glass melt circulation Effective bubble column diameter and rise velocity were extracted from the resolved simulations, compared with empirical correlations, and refit across relevant viscosity and gas flow rate conditions Explicit gas–liquid interface tracking was replaced with a single-phase momentum source term model, enabling faster steady-state CFD simulations while preserving the dominant hydrodynamic effects of bubbling New empirical correlations were developed for effective bubble column diameter and bubble rise velocity by fitting resolved simulation data across expected melter viscosity and gas flow rate ranges, providing improved inputs for the momentum source term model compared with existing literature correlations The momentum source term model reduced fluid-domain mesh size by 89% and achieved an 8.4× computational speedup relative to resolved bubbling simulations The validated momentum source term approach enables prediction of process-relevant heat transfer behavior in the integrated melter model, including heat transfer from the molten glass to the cold cap, plenum, refractory walls, and surrounding structural regions under varying melter operating conditions

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2013

Tank 48 Chemical Destruction - 13237

Small tank copper-catalyzed peroxide oxidation (CCPO) is a potentially viable technology to facilitate the destruction of tetraphenylborate (TPB) organic solids contained within the Tank 48H waste at the Savannah River Site (SRS). A maturation strategy was created that identified a number of near-term development activities required to determine the viability of the CCPO process, and subsequent disposition of the CCPO effluent. Critical activities included laboratory scale validation of the process and identification of forward transfer paths for the CCPO effluent. The technical documentation and the successful application of the CCPO process on simulated Tank 48 waste confirm that the CCPO process is a viable process for the disposition of the Tank 48 contents. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Tank 48 - Chemical Destruction

Small tank copper-catalyzed peroxide oxidation (CCPO) is a potentially viable technology to facilitate the destruction of tetraphenylborate (TPB) organic solids contained within the Tank 48H waste at the Savannah River Site (SRS). A maturation strategy was created that identified a number of near-term development activities required to determine the viability of the CCPO process, and subsequent disposition of the CCPO effluent. Critical activities included laboratory-scale validation of the process and identification of forward transfer paths for the CCPO effluent. The technical documentation and the successful application of the CCPO process on simulated Tank 48 waste confirm that the CCPO process is a viable process for the disposition of the Tank 48 contents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Evaluation of Pu Solubility in Glass for Sludge Batch 11

Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗