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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Assessment of Active Metal Removal from Pyroprocessing Salt by Electrodeposition

Active metals present in used nuclear fuel dissolve into the salt during pyroprocessing, but are not recovered during electrorefining or drawdown operations due to their chemical stability in the salt. The accumulation of 137 Cs and 90 Sr in process salt over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of removing active metals from the molten salt to enable recycle of the electrolyte (e.g., LiCl/KCl) would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Recycled Paper as a Source of Renewable Jet Fuel in the United States

Converting biomass into jet fuel involves more than the core chemical process. The overall process includes the logistics of harvesting and transporting the biomass, handling and preparing the material for processing, and processing and disposal of waste. All of these activities contribute to cost. Controlling cost involves more than developing efficient process chemistry. Choice of feedstock also has a significant impact on process economics. We consider chemical conversion of paper from municipal solid waste as a feedstock for the production of jet fuel and diesel. Paper has a significantly higher cellulose content than raw lignocellulosic biomass such as corn stover, so it requires less pretreatment to convert it into hydrocarbons than lignocellulosic biomass. Our techno-economic analysis showed that the cost of converting paper waste into jet fuel is about $1.00/gal less than jet fuel produced from corn stover. Although the cost of recycling paper into jet fuel is less than producing it from corn stover, the process is not competitive with petroleum. We estimated a minimum selling price of $3.97/gal for paper-derived jet fuel. Our sensitivity studies indicated that the biggest economic obstacle is the cost of cellulose hydrolysis. Direct hydrogenation of paper to sugar alcohols combined with increased economy of scale could make recycling paper jet fuel competitive.

09 BIOMASS FUELS↗
DOE OSTI2023

Prediction of Long-Term Geochemical Change in Bentonite Based on the Interpretative THMC Model of the FEBEX In Situ Test

Since nuclear energy is crucial in the decarbonization of the energy supply, one hurdle to remove is the handling of high-level radioactive waste (HLW). Disposal of HLW in a deep geological repository has long been deemed a viable permanent option. In the design of a deep geological repository, compacted bentonite is the most commonly proposed buffer material. Predicting the long-term chemical evolution in bentonite, which is important for the safety assessment of a repository, has been challenging because of the complex coupled processes. Models for large-scale tests and predictions based on such models have been some of the best practices for such purposes. An 18-year-long in situ test with two dismantling events provided a unique set of chemical data that allowed for studying chemical changes in bentonite. In this paper, we first developed coupled thermal, hydrological, mechanical, and chemical (THMC) models to interpret the geochemical data collected in the in situ test and then extended the THMC model to 200 years to make long-term prediction of the geochemical evolution of bentonite. The interpretive coupled THMC model shows that the geochemical profiles were strongly affected by THM processes such as evaporation/condensation, porosity change caused by swelling, permeability change, and the shape of concentration profiles for major cations were largely controlled by transport processes, but concentration levels were regulated by chemical reactions, and the profiles of some species such as pH, bicarbonate, and sulfate were dominated by these reactions. The long-term THMC model showed that heating prolongs the time that bentonite becomes fully saturated in the area close to the heater/canister; however, once the bentonite becomes fully saturated, high concentrations of ions in bentonite near the heater, which was observed in the field test, will disappear; illitization continues for 50 years but will not proceed further.

Zheng, Liange↗
DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

2020 NEPA Polychlorinated Biphenyl (PCB) Risk Assessment

The Waste Isolation Pilot Plant (WIPP) facility is a U.S. Department of Energy (DOE) operating repository 654 m below the surface in a thick salt formation in southeastern New Mexico. The DOE disposes transuranic (TRU) waste produced from atomic energy defense activities at the WIPP facility. A portion of the waste shipped to the WIPP facility contains TRU radionuclides co-mingled with polychlorinated biphenyls (PCBs), which fall under U.S. Environmental Protection Agency (EPA) regulations implementing the Toxic Substances Control Act (TSCA). This report documents the risks of PCBs co-mingled with TRU waste (hereafter designated as PCB/TRU waste) designated for disposal at the WIPP facility. This analysis is input to the National Environmental Policy Act (NEPA) assessment by the DOE Carlsbad Field Office (CBFO) for the proposed increase of the WIPP facility disposal area to include additional waste panels (but not to increase the legislated WIPP volume). This analysis is not a compliance calculation to support a certification renewal nor does it support a planned change request (PCR) or planned change notice (PCN) to be submitted to the EPA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Improbability of Nuclear Criticality in Compacted Criticality Control Overpacks after Room Closure by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, nuclear criticality is improbable after salt creep causes compaction of criticality control overpacks (CCOs) disposed at the Waste Isolation Pilot Plant, an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality potential has been updated here because of the introduction of CCOs, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCOs during two representative conditions: (1) large salt block fall, and (2) gradual salt compaction (without brine seepage and subsequent gas generation to permit maximum room closure). Geomechanical models of rock fall demonstrate three tiers of CCOs are not greatly disrupted. Geomechanical models of gradual room closure from salt creep predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Criticality models of spheres and cylinders of 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular spacing demonstrate that an array of CCOs is not critical when surrounded by salt and magnesium oxide, provided the amount of hydrogenous material shipped in the CCO (usually water and plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

New Beginnings: Optimising Waste and End-state at UK Nuclear Sites to Enable Environmental Permit surrender (the GRR) - 20193

The decommissioning of nuclear sites involves the management of large volumes of radioactive waste, other conventional waste, and the clean-up of land and groundwater contamination. In the UK, the disposal of radioactive waste from nuclear sites is regulated by the Environment Agency (England), Natural Resources Wales (Wales) and the Scottish Environment Protection Agency (Scotland). All three environment agencies seek to make sure that the generation and disposal of radioactive waste is optimised: ensuring impacts to people are 'as low as reasonably achievable' (ALARA). With a large proportion of UK nuclear sites embarking on decommissioning, the environment agencies have collaboratively developed new guidance: the 'Management of radioactive waste from the decommissioning of nuclear sites: Guidance on the Requirements for Release from radioactive substances regulation' (referred to as the GRR)[1]. The environment agencies worked in liaison with the Office for Nuclear Regulation (ONR), the UK nuclear safety regulator, and the UK government body responsible, the Nuclear Decommissioning Authority (NDA). Prior to finalising the guidance, a 12 month trial was completed at three of the NDA's nuclear sites, enabling the environment agencies to test their draft GRR guidance against 'real world scenarios' and the challenge of incorporating its requirements into decommissioning plans. Experience gained during the trial was incorporated into the current version of the GRR, published in July 2018. The GRR represents the 'new beginning' of this paper's title because for the first time, the UK environmental regulators have expressed in detail their requirements for surrender of environmental permits for nuclear sites. The GRR provides a framework for a UK nuclear site to plan, assess and justify its management of radioactive waste and deliver through decommissioning, an optimal, environmentally safe and sustainable end-state. This paper will describe the requirements and technical standards within the GRR, the establishment of defensible clean-up criteria and the production of a Waste Management Plan (WMP), supported by a Site-Wide Environmental Safety Case (SWESC). It will describe the experience gleaned by operators and regulators during the trial and it will highlight current work by regulators to implement the new approach proportionately. Finally, this paper will look forward and outline planned amendments to legislation in the UK affecting nuclear sites that should enable the full benefits of the GRR to be realised in the future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

FY21 NNSA-IAEC Science Area V, Environmental ISR, Waste Management and Subsurface Science (Final Report)

The FY21 NNSA NA-22 Final Report represents the progress achieved over the past year (FY21) by the research team for Topic Area 3, Waste Management & Subsurface Science (WM3), and Topic Area 7, Radiation and Thermal Effects on Bituminous Rocks (WM7). WM3 and WM7 are part of Science Area V (Subsurface Science and Waste Management) in the NNSA-IAEC Science and Technology Working Group. The project was initiated four years ago (2017) with an MOU agreement between the U.S. National Nuclear Security Administration (NNSA) and the Israel Atomic Energy Commission (IAEC) to evaluate the feasibility of geological subsurface disposal of radioactive waste in Israel. The core WM3 and WM7 teams are comprised of scientists and engineers from Los Alamos National Laboratory, the Geologic Survey of Israel (GSI), and the Nuclear Research Center - Negev (NRCN), with a close collaboration to the teams from Sandia National Laboratory (SNL) and Lawrence Livermore Laboratory (LLNL)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗