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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel

Epsilon metal (ε-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000°C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Non-Equilibrium Actinide Radiation Chemistry and the Nuclear Fuel Cycle

Invited John and Naomi Fackler Lectureship in Chemistry and English seminar at Valparaiso University, IN, USA. Actinides are inherently unstable elements that frequently coexist with other radioisotopes, generating intense ionizing radiation fields that drive the formation of non-equilibrium oxidation states. These transient species exert a profound mechanistic influence on the radiation response of actinide-containing systems due to their unique redox chemistry. Despite their importance, they remain poorly understood, yet such insight is essential for advancing actinide science and accurately predicting radiation-driven behavior. Actinide separations—critical for nuclear energy technologies, strategic deterrence, space exploration, and nuclear medicine—depend on precise control of actinide oxidation states to recover targeted elements from complex matrices such as used nuclear fuel. However, during these processes, actinides, their coordination complexes, and the separation media are all exposed to intense, multicomponent (alpha, beta, gamma, etc.) radiation fields that can alter process efficiency, selectivity, and chemical stability. Understanding, controlling, and mitigating radiation-induced reactions is therefore key to innovating and optimizing next-generation separation technologies. This seminar will provide an overview of the nuclear fuel cycle and non-equilibrium actinide radiation chemistry in the context of recovering actinides from used nuclear fuel, with a particular emphasis on direct-dissolution–based reprocessing strategies. We will explore time-resolved electron pulse radiolysis and gamma dose accumulation studies to elucidate the molecular-level roles of radiation-driven, non-equilibrium actinide species in process performance and in the radiolytic stability of organic ligands used for actinide recovery. These insights offer new pathways for designing advanced separation methods and next-generation solvent systems, with broad implications for the future of the nuclear fuel cycle.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2018

Criticality concerns of a group actinide co-crystallization separations approach to used nuclear fuel recycling

We report the co-crystallization of hexavalent actinides with uranyl nitrate hexahydrate has been established as a viable approach to achieving group actinide separation in used nuclear fuel recycle while avoiding the difficulties associated with the solvent-extraction process altogether. A preliminary neutronics safety analysis has been performed to alleviate criticality concerns associated with this group hexavalent actinide co-crystallization approach. The basic uranyl nitrate hexahydrate material compositions used are representative of used fuel found at commercial pressurized water reactors all around the U.S. This material is subsequently doped with various percentages of either typical reactor grade plutonium or with the total transuranic vector, replacing the uranium atoms in the crystal lattice structure. Infinite and effective neutron multiplication factors were calculated using the state-of-the-art Monte Carlo neutron transport code MCNP in a variety of plausible material configurations, assuming an acceptable safety limit for criticality to be an effective neutron multiplication factor less than 0.95. The results indicate that the crystal structures form a very subcritical intrinsic geometry even with six water molecules being incorporated into the crystal structure. In conclusion, the results of this research suggest that significant opportunity exists for the safe transfer of the group actinide separation approach from a laboratory to an operational scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Reconstituted asbestos matrix for fuel cells

Method is described for reprocessing commercially available asbestos matrix stock to yield greater porosity and bubble pressure (due to increased surface tension), improved homogeneity, and greater uniformity.

Mcbryar, H.↗
DOE OSTI2017

Benefits from developments in the field of Decommissioning for Fukushima Daiichi fuel debris retrieval: Remote-Controlled Laser Cutting Process - 17110

Among the numerous challenges towards the decommissioning of Fukushima Dai-ichi reactors, removing fuel debris is certainly one of the tougher ones, requiring several technical developments. One of them, related to laser cutting of fuel debris is being developed by a French consortium 'ONET Technologies, CEA and IRSN'. Remote controlled laser cutting was implemented for the first time in a real decommissioning workshop in 2015, in UP1 reprocessing plant. In addition to laser cutting development in air and underwater, key aspects of fuel debris cutting of representative of corium debris simulants was performed to test the quantities of aerosols, gas and dusts released during the debris laser cutting. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Cold Crucible Induction Melter (CCIM) Testing on Glass Ceramic and Ceramic Waste Forms

The U.S. Department of Energy Office of Nuclear Energy’s Fuel Cycle Research and Development (FCR&D) program is developing advanced high level waste (HLW) forms for the immobilization of the long-lived radionuclides resulting from future potential reprocessing of the used nuclear fuel from commercial reactors. One of the program’s goals is to have orders of magnitude improvements in durability (compared to the reference borosilicate glass waste form for HLW), to facilitate lower costs including less expensive and complex processing, to lower storage costs and to provide for final disposal of the waste in a deep geological repository. Both glass ceramic and ceramic waste forms are being developed and evaluated as possible improvements to borosilicate glass to achieve these objectives.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Material Balance Area Design for the Transuranic Fuel Cycle Employing High Temperature Gas Cooled Reactors

A major impetus has been to ensure safety in nuclear power plant systems from the beginning of its development to the current period of nuclear renaissance. As one of the next generation reactors with a high level of nuclear safety, high temperature gas cooled reactor (HTGR) has attracted large attention because of the possibility to shut down the reactor safely and remove decay heat without the use of active cooling. A deep-burn concept of fuel utilization is proposed and studied for HTGRs to alleviate plutonium proliferation concern. The idea is to utilize the plutonium extracted by reprocessing the light water reactor (LWR) spent fuel and discharge it only at higher fuel burn-ups of up to 90 atom %. In addition HTGRs have been studied to utilize transuranic (TRU) fuel. A large amount of special nuclear material (SNM) will be used in the HTGR advanced fuel cycle. The inherent proliferation resistance and material attractiveness of the fresh fuel and discharged spent fuel have been discussed against the diversion from peaceful to military uses. On the other hand safeguards of SNM is also important to prevent nuclear proliferation under the Non-Proliferation Treaty between the state and International Atomic Energy Agency (IAEA). Facility operators are responsible to design material balance area (MBA) and provide enough information on material balance of SNMs for the safeguards conclusion. The material unaccounted for (MUF) and measurement uncertainty in MUF (σ{sub MUF}) represent accuracy of material balance. The MBAs have to be designed and established to achieve enough confidence level on MUF and σ{sub MUF}. The MBA design is challenging for nuclear facilities managing large amount of bulk (instead of items) SNM because of potentially large MUF and σ{sub MUF}. A concept and guidance of MBA design for uranium fueled HTGRs has been discussed. In this study, a MBA design is presented for the transuranic fuel cycle employing HTGRs. The MUF and the σ{sub MUF} is calculated and evaluated using international target values (ITVs) provided by the IAEA. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Grouped actinide separation in advanced nuclear fuel cycles

Aiming at cleaner waste streams (containing only the short-lived fission products) a partitioning and transmutation (P-T) scheme can significantly reduce the quantities of long-lived radionuclides consigned to waste. Many issues and options are being discussed and studied at present in view of selecting the optimal route. The choice is between individual treatment of the relevant elements and a grouped treatment of all actinides together. In the European Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), grouped separation options derived from an aqueous extraction or from a dry pyroprocessing route were extensively investigated. Successful demonstration tests for both systems have been carried out in the frame of this project. The aqueous process called GANEX (Grouped Actinide Extraction) is composed of 2 cycles, a first one to recover the major part of U followed by a co-extraction of Np, Pu, Am, and Cm altogether. The pyro-reprocessing primarily applicable to metallic fuels such as the U-Pu-Zr alloy originally developed by the Argonne National Laboratory (US) in the mid 1980s, has also been applied to the METAPHIX fuels containing up to 5% of minor actinides and 5% of lanthanides (e.g. U{sub 60}Pu{sub 20}-Zr{sub 10}Am{sub 2}Nd{sub 3.5}Y{sub 0.5}Ce{sub 0.5}Gd{sub 0.5}). A grouped actinide separation has been successfully carried out by electrorefining on solid Al cathodes. At present the recovery of the actinides from the alloy formed with Al upon electrodeposition is under investigation, because an efficient P-T cycle requires multiple re-fabrication and re-irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

An advanced hybrid reprocessing system based on UF{sub 6} volatilization and chromatographic separation

To recover U, Pu, MA (Np, Am, Cm) and some specific fission products FPs (Cs, Sr, Tc, etc.) from various spent nuclear fuels (LWR/FBR: Oxide, Metal Fuels), we are studying an advanced hybrid reprocessing system based on UF6 volatilization (Pyro) and chromatographic separation (Aqueous). Spent fuels are de-cladded by means of thermal and mechanical methods and then applied to the fluorination/volatilization process, which selectively recovers the most amount of U. Then, the remained fuel components are converted to oxides and dissolved by HNO{sub 3} solution. Compared to U, since Pu, MA and FPs are significantly less abundant in spent fuels, the scale of the aqueous separation process could become reasonably small and result in less waste. For the chromatographic separation processes, we have prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the actinides and FP elements from the fuel dissolved solution. In this work, adsorption and separation behavior of representative actinides and FP elements was studied. Small scale separation tests using simulated and genuine fuel dissolved solutions were carried out to verify the feasibility of the proposed process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Direct disposal of spent fuel: developing solutions tailored to Japan

With the past Government policy of 100% reprocessing in Japan now open to discussion, options for direct disposal of spent fuel (SF) are now being considered in Japan. The need to move rapidly ahead in developing spent fuel management concepts is closely related to the ongoing debate on the future of nuclear power in Japan and the desire to understand the true costs of the entire life cycle of different options. Different scenarios for future nuclear power - and associated decisions on extent of reprocessing - will give rise to quite different inventories of SF with different disposal challenges. Although much work has been carried out spent fuel disposal within other national programmes, the potential for mining the international knowledge base is limited by the boundary conditions for disposal in Japan. Indeed, with a volunteer approach to siting, no major salt deposits and few undisturbed sediments, high tectonic activity, relatively corrosive groundwater and no deserts, it is evident that a tailored solution is needed. Nevertheless, valuable lessons can be learned from projects carried out worldwide, if focus is placed on basic principles rather than implementation details. (authors)

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

Progress and Technology Assessment of Melt-Cast Ceramic Waste Forms

The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.

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DOE OSTI2018

Delivering Business Benefit to UK Decommissioning through Collaborative Innovation - Approach and Technologies - 18539

The UK's nuclear energy programme, dating from the post-war years, has left a mixed legacy to the country: numerous prototype reactors, fuel-manufacturing plants, research centres, reprocessing plants and 11 power stations. The Sellafield site in west Cumbria houses more than 200 nuclear facilities and 1,000 buildings, making it one of the world's most complex environmental decommissioning challenges. Many 'never-done-before' projects are under way, which require substantial levels of R and D and innovative engineering approaches. One of the Nuclear Decommissioning Authority's (NDA) responsibilities is to ensure that the right level of R and D is carried out to enable progress on delivering the full decommissioning programme. Since 2010, NDA has been collaborating with Innovate UK, the UK government's innovation agency, to engage the supply chain and develop innovative technologies and services to support the NDA's mission and the growing international nuclear decommissioning market. The paper will provide an overview of the approach (e.g. open calls, briefing events, independent assessment, technology showcases) and the associated benefits (e.g. attracting new ideas, end-user engagement, increased supply chain collaboration). The paper also highlights selected technologies developed through the process. (authors)

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

West Valley Demonstration Project (WVDP) Annual Site Environmental Report (ASER) for Calendar Year 2024

The report, prepared for the U.S. Department of Energy West Valley Demonstration Project office (DOE-WVDP), summarizes the environmental protection program at the WVDP for calendar year (CY) 2024. Monitoring and surveillance of the facilities used by the DOE are conducted to verify protection of public health and safety and the environment. The report is a key component of DOE’s effort to keep the public informed of environmental conditions at the WVDP. The quality assurance protocols applied to the environmental monitoring program ensure the validity and accuracy of the monitoring data. In addition to demonstrating compliance with environmental laws, regulations, and directives, evaluation of data collected in 2024 continued to indicate that WVDP activities pose no threat to public health or safety, or to the environment.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2026

West Valley Demonstration Project (WVDP) Annual Site Environmental Report (ASER) for Calendar Year 2025

The report, prepared for the U.S. Department of Energy West Valley Demonstration Project office (DOE-WVDP), summarizes the environmental protection program at the WVDP for calendar year (CY) 2025. Monitoring and surveillance of the facilities used by the DOE are conducted to verify protection of public health and safety and the environment. The report is a key component of the DOE’s effort to keep the public informed of environmental conditions at the WVDP. The quality assurance protocols applied to the environmental monitoring program ensure the validity and accuracy of the monitoring data. In addition to demonstrating compliance with environmental laws, regulations, and directives, the evaluation of data collected in 2025 continued to indicate that WVDP activities posed no threat to public health or safety, or to the environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 1, January-March 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Coplnger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.

05 NUCLEAR FUELS↗
DOE OSTI2017

Feasibility of a Breed-and-Burn Molten Salt Reactor

This work addresses the feasibility and characteristics of the breed and burn (B and B) cycle for Molten Salt Reactors. In the general B and B cycle, natural Uranium or Thorium is used as fuel for a reactor which breeds a sufficient amount to maintain criticality at equilibrium without reprocessing. Interest in applying this scheme to an MSR has developed recently For the MSR B and B scheme considered here, fresh salt containing fertile material is continually fed into the core while salt with the composition of the homogeneous fuel is expelled without reprocessing. Through an analysis of the equilibrium conditions and startup requirements, the feasibility of such a cycle operating with various salts is assessed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Progress Towards Treating the Sludges Produced During Spent Fuel Processing - 19546

As a consequence of over 50 years of Magnox reprocessing, Sellafield site is in possession of significant quantities of Magnox sludge. This sludge material is the by-product of corrosion of the magnesium based Magnox metal that clads the Magnox fuel elements. Retrievals and processing of this material poses a significant challenge as corrosion of Magnox metal has occurred under a variety of conditions and over a significant period of time, in a variety of locations. Devising processes to treat heterogeneous materials offers a significant challenge to identify technologies with robust envelopes of acceptance, to effectively operate together. This is further compounded by difficulties defining test programmes and materials for testing that will provide assurance that nonradioactive testing is suitably representative or bounding. An overview of the significant body of work undertaken to process Sellafield's Magnox sludge is provided, from the thousands of commercially sensitive reports generated over the past five decades. This covers the sampling, characterisation, strategy and process evolution, technology development and test materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗