Microfluidic Spectroscopic Sensor for Nuclear Fuel Reprocessing Solutions Phase I Final Report
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46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
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Recent Advances in Used Nuclear Fuel Reprocessing Off-Gas Treatment
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Jubin, Robert↗
DEVELOPMENT OF A CONTINUOUS COST EFFECTIVE KR XE CAPTURE PROCESS FOR NUCLEAR FUELS REPROCESSING
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Zdyrko, Bogdan↗
Spent Nuclear Fuel and Reprocessing Waste Inventory: Spent Fuel and Waste Disposition
This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers as of the end of calendar year 2021. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios, based on information available and assumptions made at the time the scenarios were developed in the spring of 2022. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities
The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. As a result, the experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel
The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Monitoring Aqueous Reprocessing Systems for Detection of Facility Misuse
Traditional approaches to material control and accounting at used nuclear fuel reprocessing facilities relies in large part on laboratory-based destructive assay of periodic samples taken from key stages. These techniques are highly accurate, but limited inspection frequency may limit the size and operation of a reprocessing facility. Online monitoring techniques using inline detectors can be applied to fuel reprocessing systems to improve the situational awareness as the facility is operating to provide additional indicators of the evolving state of the facility between normal inspection periods, potentially alerting regulators to the need for additional inspection. The Multi-Isotope Process (MIP) Monitor combines medium-resolution gamma detectors with statistical data analysis techniques to monitor reprocessing systems for anomalous conditions resulting from process control issues or malfeasance. Gamma spectra collected from the H-Canyon reprocessing facility were modeled using principal component analysis to demonstrate the efficacy of the MIP Monitor for detecting notional diversion scenarios.
safeguards, nuclear fuel reprocessing, gamma spect↗
Nuclear Fuels and Reprocessing Technologies: A U.S. Perspective
Reprocessing and/or waste management issues are of concern to the “back end” of the nuclear fuel cycle. Of course, there are a great many “nuclear fuel cycle” scenarios to consider; if not in practice, then at least in theory. The simplest conceptually is the “once through” fuel cycle in which the spent fuel is discarded. The more complex fuel cycle scenarios involve reprocessing spent nuclear fuels and a family of nuclear reactor technologies to accommodate burning and breeding for various military and commercial needs. Therefore, the selection of a specific “fuel cycle” is what ultimately imposes the engineering requirements of the reprocessing and waste management technologies. No one part is independent of the other parts in a fuel cycle flowsheet; all parts are fully integrated. This paper presents a summary of radiochemical processes, nuclear reactor technologies, nuclear fuel types, and the reprocessing technologies that serve the different nuclear fuel types. Comprehending how this series of topics are related to each other is a prerequisite to understanding the requirements of any reprocessing strategy. The summary materials presented here are selective, as opposed to comprehensive. More detailed information on any one subject can be found in the reference materials.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Enhancements to a dissolver burn up credit safety case to reduce further GD use in the thermal oxide reprocessing plant (THORP), Sellafield - 14229
THORP reprocesses fuel from Britain's Advanced Gas-cooled Reactors (AGRs) and fuel from Light Water Reactors (LWRs). The THORP dissolvers use nitric acid poisoned with Gadolinium (Gd) to ensure criticality safety. The dissolver criticality safety case uses a burnup credit approach to determine a minimum required Gd concentration. The burn up credit approach followed originates from a time when fuel receipts for THORP were ongoing and hence models a deliberately conservative and artificial fuel type to cover all potential fuel receipts. External fuel receipts have now ceased and detailed information on post irradiation U and Pu isotopics is available for all fuel scheduled for reprocessing. Work has therefore been undertaken to determine whether the burn up credit safety case could be revised to be more representative of the fuel actually remaining for reprocessing and hence to reduce further the required Gd concentration for the dissolvers. This would ultimately reduce the quantity of high active liquid waste requiring treatment and storage. The great majority of fuel remaining for reprocessing in THORP is AGR fuel, for which the typical burnup is known to be much higher than was previously assumed in the safety case. Pu and U isotopics for irradiated AGR fuel were examined with the fuel inventory code FISPIN. Modelling with the code MONK then showed that the Gd requirement for irradiated AGR fuel of a given burnup was much lower than in the previous safety case due to the change in Pu:U ratio and Pu{sup 240} in Pu for AGR fuel compared with the previous conservative assumptions. The burnup of all remaining LWR fuel was found to be high and hence lower dissolver Gd concentrations could also be tolerated for reprocessing this fuel. The work has concluded that the THORP dissolver Gd concentration can be halved for all remaining LWR and AGR fuel campaigns with a consequent cost saving of ∼ UK pound 4 m in Gd and high active waste processing time. Implementation of this improvement is currently ongoing on the plant. (author)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Multivariate analysis of gamma spectra to characterize used nuclear fuel
Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.
Multivariate Analysis for Characterization of Used↗
Magnox Reprocessing TDN Reactors: Utilising 3D Printing and X-Ray Imaging to Re-Design and Test Fluidising Air Nozzles - 16402
One of the key final steps in reprocessing spent nuclear fuel is the conversion of highly concentrated uranyl nitrate hexahydrate liquor [> 1000 gU/litre] into uranium trioxide (UO{sub 3}) powder by thermal decomposition using Thermal Denitration (TDN) reactors. The TDN reactors in the Magnesium non-oxidising (Magnox) fuel reprocessing plant are high aspect ratio fluidising beds, which suffer from poor fluidization and UO{sub 3} powder blockages. It is widely understood that these issues arise from the inadequate design of the fluidising air distributors (nozzles), which are located around the base of the TDN reactors. A non-active test rig was constructed at the National Nuclear Laboratory (NNL) experimental facility in Workington, Cumbria (UK) which used compressed air at ambient temperature to assess the performance of the current fluidising air nozzle designs (datum) against a range of new nozzle designs. Full-scale nozzle designs were manufactured using a 3-Dimensional (3D) printer and tested to obtain data of both quantitative and qualitative nature. NNL then collaborated with the Chemical Engineering Department at University College London (UCL) to further assess the fluidising potential of the new nozzle designs. Images of inside the fluidized bed were taken utilising a unique X-Ray facility. The results from this study will allow new nozzle designs to be recommended, manufactured and installed, which will ensure improved fluidization of the TDN reactor bed and prevent blockages with UO{sub 3} powder. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Aqueous Reprocessing of Used Nuclear Fuel
Aqueous reprocessing of used nuclear fuel has a long history of successful operation worldwide. The history of aqueous reprocessing and a description of current and past operating aqueous reprocessing facilities are provided and the main types of solvent extraction equipment utilized in the nuclear industry are described. While the PUREX process is well established for separating U and Pu from used nuclear fuel, numerous variants to the PUREX process are being developed and are described. Finally, advanced separation processes in which minor actinides are separated from the used nuclear fuel are a key research focus worldwide. These processes and the current state of development are discussed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Liquid Wastes Calcination: The High-Level French Experience - 17184
Calcination is an efficient process for volume reduction and stabilization of nuclear liquid wastes. It is often the first step of waste containment processes such as vitrification or grouting. The CEA (French Alternative Energies and Atomic Energy Commission) and AREVA have acquired a high-level experience in the field of calcination through more than forty years of R and D and industrial operation. During this period, a broad range of liquid wastes have been studied and treated. R and D work and associated engineering studies allowed defining the design and sizing of the equipment and the process control parameters to be applied for the calcination of all kinds of solutions. The calcination process has been qualified for the following effluents amongst others: - Liquid waste from UOx fuel reprocessing; - Liquid waste from UMo fuel reprocessing (high molybdenum and phosphorus contents); - Hanford HLW; - Liquid waste from D and D operation (sodium bearing waste). This paper presents several studies led by the Joint Vitrification Laboratory (L.C.V), a common research laboratory between CEA and AREVA in charge of qualifying new processes and matrices for waste containment. These studies allowed defining the calcination rules, in particular for very hard to process solutions such as effluents with high boron, molybdenum and sodium contents. The feedback from industrial operation of the calcination process with fission products and effluents from D and D operations, highly loaded with sodium, is also presented. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Spent Nuclear Fuel and Reprocessing Waste Inventory
This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Spent Nuclear Fuel and Reprocessing Waste Inventory
This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers as of the end of calendar year 2024. Actual quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios, based on information available and assumptions made at the time the scenarios were developed. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Spent Nuclear Fuel and Reprocessing Waste Inventory
This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).
Peters, Shan↗
Spent Nuclear Fuel and Reprocessing Waste Inventory
This report provides information on the inventory of commercial spent fuel (SNF) and high-level radioactive waste in the United States, as well as non-commercial SNF and reprocessing waste in the U.S. Department of Energy (DOE) complex. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future commercial nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Synthesis and Characterization of 5- and 6- Coordinated Alkali Pertechnetates
The local chemistry of technetium-99 ( 99 Tc) in oxide glasses is important for understanding the incorporation and long-term release of Tc from nuclear waste glasses, both those for legacy defense wastes and fuel reprocessing wastes. Tc preferably forms Tc(VII), Tc(IV), or Tc(0) in glass, depending on the level of reduction of the melt. Tc(VII) in oxide glasses is normally assumed to be isolated pertechnetate TcO 4 - anions surrounded by alkali, but can occasionally precipitate as alkali pertechnetate salts such as KTcO 4 and NaTcO 4 when Tc concentration is high. In these cases, Tc(VII) is 4-coordinated by oxygen. A reinvestigation of the chemistry of alkali-technetium-oxides formed under oxidizing conditions and at temperatures used to prepare nuclear waste glasses showed that higher coordinated alkali Tc(VII) oxide species had been reported, including those with the TcO 5 - and TcO 6 - anions. The chemistry of alkali Tc(VII) and other alkali-Tc-oxides is reviewed, along with relevant synthesis conditions. Additionally, we report attempts to make 5- and 6-coordinate pertechnetate compounds of K, Na, and Li, i.e. TcO 5 - and TcO 6 -. It was found that higher coordinated species are very sensitive to water, and easily decompose into their respective pertechnetates. It was difficult to obtain pure compounds, but mixtures of the pertechnetate and other phase(s) were frequently found, as evidenced by x-ray absorption spectroscopy (XAS), neutron diffraction (ND), and Raman spectroscopy. Low temperature electron paramagnetic resonance (EPR) measurements showed the possibility of Tc(IV) and Tc(VI) in Na 3 TcO 5 and Na 5 TcO 6 compounds. Here, it was hypothesized that the smaller counter cation would result in more stable pertechnetates. To confirm the synthesis method, LiReO 4 and Li 5 ReO 6 were prepared, and their Raman spectra match those in the literature. Subsequently, the Tc versions LiTcO 4 and Li 5 TcO 6 were synthesized and characterized by ND, Raman spectroscopy, XANES, and EXAFS. The Li 5 TcO 6 was a marginally stable compound that appears to have the same structure as that known for Li 5 ReO 6 . Implications of the experimental work on stability of alkali technetate compounds and possible role in the volatilization of Tc are discussed.
36 MATERIALS SCIENCE↗