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At least 1,585 records · Page 88
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

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

Generic Post-Closure Safety Assessment for Deep Borehole Disposal of Heat-Generating Waste - 20451

The deep borehole disposal (DBD) concept for high-activity waste consists of drilling a large-diameter borehole into crystalline basement rock to a depth of about 5,000 m, placing waste packages in the lower portion of the borehole - the waste emplacement zone - and sealing and plugging the upper portion of the borehole with a combination of bentonite, cement plugs, and sand or crushed rock ballast. The waste emplacement zone is several times deeper than typical mined repositories and is well below the typical maximum depth of fresh groundwater resources. Recent work describes a generic reference case for deep borehole disposal of Cs and Sr capsules stored at the Hanford Site. Two scenarios are described: an undisturbed (nominal) scenario, and disturbed scenario, in which a waste package is stuck adjacent to a transmissive fracture zone in crystalline basement. Simulations of the processes affecting radionuclide release and transport in the two scenarios provide the basis for quantitative post-closure performance assessment (PA) of the reference case. This paper presents a probabilistic PA of the stuck waste package scenario, i.e. uncertainties affecting radionuclide mobilization and transport are propagated through 200 realizations of a predictive model to quantify uncertainties in predicted radionuclide concentrations within the borehole and the fracture zone. Simulations predict no radionuclide transport in the borehole to distances greater than 100 m above the stuck waste package. An assumed lateral pressure gradient results in predictions of limited radionuclide transport in the fracture zone at a lateral distance of 200 m from the stuck waste package. These results complement previous results for the nominal scenario and contribute to the development of a generic safety case for DBD. (authors)

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

Post-irradiation Experiment on IG-110 and A3-3 Graphites for HTGR - 18494

The first irradiation test of A3-3 graphite matrix and IG-110 graphite were performed in Hanaro during the period of August 2013 through March 2014. The dimensional and weight changes, microstructure, and thermal diffusivity were investigated in different directions of the specimen after the irradiation. Interestingly, the density of the A3-3 matrix was increased by 9.4% due to the weight increase after the irradiation. Moreover, the thermal diffusivity in the radial direction of the A3-3 matrix is almost 3-times higher than that of the axial direction owing to the radially oriented graphite texture shown in the microstructure. Meanwhile, 10-20 {mu}m sized irradiation induced pores are observed in the microstructure of IG-110. The thermal diffusivity measurement of pre- and post-irradiated IG-110 results in a 68% irradiation-dominant decrease of thermal conductivity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Diversity for safety and security of NPP I and C: Post NUREG/CR 7007 stage - 306

The paper summarizes authors' experience on diversity aspects of NPP I and C systems. It also discusses diversity analysis aspects according to NUREG/CR 7007 and other relevant standards. Diversity implementation challenges related to regulatory authorities, developers, auditors and operation engineers and managers are formulated and discussed. The paper presents a case study and developed tool to support diversity analysis and assessment using post NUREG/CR 7007 approach. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Statistical Analyses of Post Irradiation Experiments of the BWR Forsmark

Computer codes which are to be used in nuclear safety research need to be validated prior to their application. For depletion codes, this validation can be done using data from post irradiation experiments (PIE). These experiments consist of a radiochemical analysis of small parts of a fuel rod (FR), which has been irradiated in a nuclear power plant as part of a fuel assembly (FA). This analysis allows the deduction of the isotopic composition of nuclides important for the desired application case. In addition the operation history at the sample location needs to be known and reported with the analysis results. In our work, we perform uncertainty and sensitivity analyses of a model based on a sample from the boiling water reactor (BWR) Forsmark-3 in order to deduce the influence of uncertain system parameters on the nuclide composition of the depleted fuel. In a second step we compare the resulting uncertainties of the nuclide densities with the deviation between their calculated and experimental values. The goal is to obtain quantitative uncertainty data which can be accounted for in a validation procedure. We use the Monte Carlo sampling approach, i.e. repeating the model calculation using statistically varied sets of the uncertain input parameters and analyzing the distributions of the result values statistically. BWR depletion calculations are especially demanding due to the increased complexity and heterogeneity compared to pressurized water reactors (PWR): extensively variable enrichments, complex FA geometries, partial length FRs, strong axially varying moderator density and the common use of gadolinium as burnable absorber in some of the FRs. However, the precise and correct calculation of BWR-depletion is extremely important in view of a possible application in burnup credit for BWR FAs. Therein credit is taken for the reduced reactivity of spent fuel due to the depletion of {sup 235}U for transport and intermediate or final storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

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

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Mitigation Methods for Piperazine Oxidation in Post-Combustion Carbon Capture

Piperazine is a promising second-generation solvent for amine scrubbing in post-combustion CO2 capture. However, the oxidative degradation of PZ can cause environmental problems and economic loss. This work presents the effects of two mitigation methods: carbon treating and N2 sparging, on the PZ oxidation in long-term operations. The species and their respective quantities adsorbed by the activated carbon were tested in a bench-scale device. The carbon was then tested in the High Temperature Oxidation Reactor (HTOR), where the solvent oxidizes at a reasonably fast rate. Pilot plant campaigns were also performed at the UT Austin SRP and the National Carbon Capture Center (NCCC) and the effects were verified. When dissolved Fe is removed by carbon treating or other methods, available soluble Fe in the system dissolves and replaces the dissolved Fe. Therefore, all available Fe and ligands need to be removed for the mitigation to be effective. The sources of soluble Fe include fly ash and the corrosion of stainless steel, and the ligands are degradation products. 8 All PZ degraded solvents have two absorbance peaks at 320 nm and 538 nm. The 320 nm peak is caused by dissolved metals, especially Fe, complexed by degradation products. The 320 nm peak is related to the amine degradation level and can be used as a simple and efficient method to estimate the amine degradation rate. The pilot plant data suggest that NO2 can oxidize PZ significantly, possibly through radical reactions. 0.01 mmol/kg-hr absorption of NO2 increased the PZ oxidation rate from 1.2 mmol/kg-hr to 2.5 mmol/kg-hr.

Yuying, Wu↗
DOE OSTI2024

UN-SiC TRISO Post Irradiation Examination Developmental Work

As part of efforts to strengthen INL?s post irradiation analysis capabilities with non-Advanced Gas Reactor (AGR) tristructural isotropic (TRISO) fuels, two developmental activities were conducted. The first activity was to determine how best to analyze uranium nitride TRISO fuel kernels using electron probe microanalysis, while the second activity focused on developing a method to deconsolidate TRISO fuel particles that have been encased in a silicon carbide matrix. Because these two activities were unrelated, they have been presented separately in this report. Initial EPMA analyses showed nitrogen contents that far exceeded the concentration expected for UN--a line compound. Further examination showed that current literature values for the mass absorption coefficient (MAC) for the N ka X-ray absorbed by U ranged from approximately 1600 to 9500, with most values tending toward 9500. Measuring UN with five different progressively increasing accelerating voltages followed by using the modeling program xMAC suggests the actual MAC is approximately 2115. Additional MAC modifications were required to produce reasonable analytical results. Because of the inaccuracies of necessary MAC coefficients, UN analysis via scanning electron microscopy (SEM) is likely to produce inaccurate results. This is because SEM software does not typically allow the user to alter MACs. Tests have been performed to examine the feasibility of an electrochemical technique to liberate irradiated TRISO fuel from a SiC matrix without damaging the outer pyrolytic carbon layer of the fuel particle. The method is performed by electrochemically exposing the SiC to magnesium metal forming Mg2Si and C. Following exposure, the small SiC samples showed slight mass increases with no evidence of conversion to Mg2Si and C nor obvious degradation of the SiC samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Thermal-hydraulics capabilities in predicting the post-CHF conditions

Developing research in a time-at-temperature limit for fuel failure requires assessing both the material modeling and the thermal-hydraulic modeling capabilities. Predictions of post-CHF conditions have areas for development leveraging experiments, code coupling, and general model improvement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS

Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.

Harp, Jason↗
DOE OSTI2024

Radial Deconsolidation Leach-Burn-Leach of Five AGR-3/4 TRISO Fuel Compacts after Post-Irradiation Heating Tests

Five AGR-3/4 fuel compacts were previously heated in the Fuel Accident Condition Simulator (FACS) furnace. Three of these compacts were also reirradiated in the Neutron Radiography (NRAD) reactor before the FACS tests. One of these was heated to 1200°C, two were heated to 1400°C, one was heated to 1600°C, and one was held for a period of time at 1600 and 1700°C. After these FACS tests, the compacts were subjected to destructive post-irradiation examination (PIE) via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantities of nuclides of interest to give volumetric concentrations as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared with their sibling compacts within the same capsule (similar irradiation conditions) that went through RDLBL in the as-irradiated state. These data will be used to refine the fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AFIP-7 Post-irradiation Examination Summary Report

The following report contains the results and conclusions from post irradiation examinations performed on the fuel plates irradiated as part of the AFIP-7 experiment, the first test of base monolithic fuel in a fuel assembly consisting of curved, constrained fuel plates. These examinations include visual examination; eddy-current testing to measure oxide growth; profilometry to measure dimensional changes of the fuel plates; neutron radiography for evaluating fuel condition prior to sectioning; gamma scanning to provide relative measurements of burn-up and indication of fission-product relocation; analytical chemistry to benchmark the physics burn-up calculations; metallographic examination to identify microstructural changes in the fuel, interlayer and cladding; and blister threshold testing. These characterization activities were conducted to demonstrate that the fuel meets irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior. Revision 1 being reviewed for share publicly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Reirradiation and Post-reirradiation Heating Tests of Loose AGR-2 TRISO Particles and AGR-3/4 Fuel Compacts

To address the lack of data on the behavior of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.24 d) in the TRISO fuel system, loose AGR-2 particles and kernels were reirradiated in NRAD and subjected to post-reirradiation heating in the FACS furnace. This provided data on the release of fission products from exposed kernels, and the results were used to establish FACS condensation plate collection efficiencies for temperatures of 1000, 1200, and 1400°C where previously, the only collection efficiencies were for 1600°C tests. These new collection efficiencies were then applied to the NRAD-FACS tests of entire AGR-3/4 fuel compacts. Understanding the relative behavior of short-lived I-131, for example, is important because it is a significant contributor to offsite dose during postulated accidents, and Xe-133 has previously been used as a conservative indicator of I-131 behavior. These data will be used to support refinement of fission product transport models and HTGR source-term analyses.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Techno-economic assessment of emissions mitigation technologies for post-combustion CO2 capture using AMP/PZ

Minimizing the environmental impacts of amine-based post-combustion carbon capture technologies is essential for meeting environmental permitting regulations and ensuring public acceptance. Experimental test campaigns at the CO₂ capture pilot plant in Niederaussem using CESAR1 demonstrated that integrating available emission abatement technologies can significantly reduce the concentration of amines and degradation products in CO₂-depleted flue gas to below the detection limit of an infrared spectrometer. The study confirmed that proprietary dry bed technology (OEASE Aerozone™) or a second water wash can lower AMP and PZ emissions to below 1 mg/Nm³. However, to achieve very low NH₃ emissions below 2 mg/Nm³, an acid or other chemically active wash downstream of the water wash is required. A configuration with a dry bed or a double water wash results in a carbon capture cost (CCC) of 44 €/tCO₂, and a CO₂ avoided cost (CAC) of 86 €/tCO₂. A configuration with an acid wash increases the CCC to 47 €/tCO₂ and the CAC to 90 €/tCO₂ due to the amine losses in the acid waste and its treatment.

CO2 capture↗
DOE OSTI2025

Post-irradiation 1600°C Heating Test of AGR-1 UCO Fuel Kernels

Five bare kernels were selected from the U.S. Advanced Gas Reactor (AGR) 1 irradiation experiment Compact 5-3-1 to perform post-irradiation safety tests. The safety test involved heating the kernels in the Fuel Accident Condition Simulator (FACS) furnace in the inert atmosphere to a peak temperature of 1600°C and isothermally holding at this temperature for about 47 hours while collecting fission products released. This test was to assess the retention of fission products in bare kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material. The bare kernels released nearly 100% of cesium and antimony, while they were able to maintain about 30% of europium, 50% of strontium, and the majority of cerium and ruthenium. In addition, about 48% of the calculated inventory of Kr-85 released during the test, indicating that kernels were capable of retaining a considerable fraction of fission gas K-85 during irradiation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗