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At least 127 records · Page 7

Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels using the Polaris/PARCS Two-Step Approach. Vol. I: PWR Fuel

Nuclear fuel with extended enrichment (235U enrichment within 5-8 wt%) is one of the evolutionary changes that have been pursued in recent years by commercial light water reactor operators and fuel vendors to improve the fuel cycle economy and operation performance of a nuclear plant. This work assesses the performance of the Polaris/PARCS two-step approach in core physics modeling of the pressurized water reactor cores with extended enrichment fuel, referred to as “LEU+” in this report. A representative LEU+ core with a 24-month fuel cycle developed by Southern Nuclear Company (SNC) was modeled using this two-step approach. A representative LEU core with an 18-month fuel cycle was also modeled to provide a reference for the LEU+ core. As expected, significantly more burnable poison absorbers were used in the LEU+ core to accomodate its higher fuel enrichment. Nine different fuel assembly types were modeled using Polaris for each core to generate the assembly cross sections, which were then processed by GenPMAX to prepare the cross-section data for PARCS. The average specific powers of each fuel batch in each core were derived from VERA results and higher specific powers in fresh assemblies were found in the LEU core due to its less total uranium loading included in the VERA LEU model, given that the total core power was assumed to be the same for both cores. PARCS models were developed to simulate the steady-state operations of both cores. PARCS results on the LEU+ core were first compared with the VERA results for verification purpose; good agreements were seen in soluble boron and burnup distribution results, indicating that the Polaris/PARCS modeling and simulation were correctly implemented. Core physics parameters calculated by PARCS, at zero power physics tests, beginning of cycle (BOC), and end of cycle conditions (EOC), were compared between the LEU+ core and the LEU core, including soluble boron concentration, burnup distributions, assembly and pin power peaking factors, fuel temperature reactivity coefficients, moderator temperature and density reactivity coefficients, control rod worth, and shut down margin. The main differences in PARCS results between the LEU+ and the LEU cores are summarized below: 1)The critical boron concentrations were found to be much higher in the LEU+ core than in the LEU core (1582 vs. 1335 ppm for peak values). 2)Higher assembly radial power peaking factors (1.4 vs. 1.3 for peak values), 2D pin peaking factors (1.53 vs. 1.42 for peak values), and 3D pin peaking factors (1.89 vs. 1.81 for peak values) were found in the LEU+ core than in the LEU core. 3)Significantly higher reactivity coefficients of moderator temperature (and density) were found in LEU+ than in LEU.4)Significantly lower control rod worth at EOC were found in LEU+ than LEU for all but one control banks.5)Significantly lower shut down margins were found in the LEU+ core than in the LEU core. The Polaris/GenPMAX/PARCS code suite was found to be capable of modeling the LEU+ PWR core for steady-state operations and no unexpected results in core physics parameters were observed, in spite of that a) several bugs in PARCS were identified and workarounds were used; b) several features were found lacking in the current version of PARCS that would be useful for core modeling. A list of requests for bug fixes and feature upgrades for PARCS originated from this work were transmitted to the code developers. The assessments on the performance of the Polaris/PARCS two-step approach in core modeling for boiling water reactor with LEU+ fuel is ongoing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Safety Analysis for Accident-Tolerant Fuels with Increased Enrichment and Extended Burnup

The ERP research and development efforts in FY22 focused on the safety analysis of ATF with increased enrichment and extended burnup to provide scientific knowledge of the ATF fuel performance, failure mechanism and source term analysis during severe accident of the NPP. FeCrAl clad ATF was selected as basis of the reactor fuel. An optimized equilibrium cycle was developed for 24-months extended burnup operation. A 4-loop PWR model under loss-of-coolant accident (LOCA) was assessed and the behavior of the major source term from failed ATF has been analyzed. The result showed that grace to enhanced mechanical characteristics of ATF, the fuel failure starts up to 200~500 seconds later than conventional Zr clad fuel. This time gives sufficient buffer until low pressure safety injection (LPSI) of the emergency core cooling system (ECCS) restarted by the operator which reduces fuel fragmentation, relocation and dispersal (FFRD). As a consequence, the total amount of major radioactive materials released from FeCrAl clad fuel into the reactor containment was found up to three times less compare to Zr clad fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Combined TREAT-LOC and SATS LOCA Experiment Plan: Integral LOCA Experiments on High-Burnup Fuels

The Transient Reactor Test Facility (TREAT) loss-of-coolant (LOC) and high-burnup (HBu) experiment series, along with the Severe Accident Test Station (SATS) HBu experiment series, are integral LOC accident (LOCA) experiments planned under the DOE AFC program, which aim to support burnup extension needs by addressing identified R&D priorities in order to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program has been specifically designed to address knowledge gaps and opportunities identified through a detailed review of existing public knowledge on LOCA FFRD. The test program employs a unique combination of in- and out-of-pile experimental approaches and state-of-the-art facilities to provide a clear connection to the existing integral and semi-integral LOCA experiment database. The primary goal of the program is to investigate the impact of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions that have not yet been fully studied. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions. Importantly, TREAT’s unique capability will enable the first evaluation of the impact of SEH conditions on HBu fuels. The test program will emphasize the development of an improved mechanistic understanding of key experimental phenomena through independent experimental systems, development of a database to support fuel performance modeling tools and employing world-leading advanced materials characterization and in-situ diagnostics to evaluate FFRD. The results of the program will provide novel data to support modeling development and validation and will represent a significant advancement in evaluating prototypic conditions, as well as to inform the technical basis for LOCA-induced FFRD.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Advanced Multiscale Microscopy Characterization of High Burnup LWR UO 2 Before and After LOCA Testing

Previous microscopy work within the campaign has observed increased porosity coupled with subgrain formation in the mid-radial region of the fuel. However, examinations of high burnup fuel taken from multiple reactors with different operating conditions have shown that this structure can move radially inward depending on the fuel operating conditions. It is currently theorized that inter- and intragranular fission gas bubble nucleation is precursory to the grain subdivision observed in the mid-radial and central regions of the fuel. Characterization of samples after loss-of-coolant accident (LOCA) testing has revealed that these restructured regions with a high density of bubbles and subgrains pulverize during the transient test. It appears that the increase in fission gas bubbles coupled with grain subdivision renders the fuel mechanically weaker during a LOCA transient and thus susceptible to fuel fragmentation relocation and dispersal (FFRD). Work this past fiscal year has prioritized understanding this behavior by using advanced microscopy to investigate fission gas behavior in the as-irradiated and post-LOCA state. Additional work has been performed to verify the theory described above by analyzing multiple post-LOCA optical micrographs and comparing that to as-irradiated microstructural data. This document reports progress in the post-irradiation characterization of high burnup nuclear fuel with emphasis on the restructured fuel regions, particularly how microstructural features influence FFRD and fission gas release behavior in LOCA conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Computational and Experimental Investigation of Thermal-Mechanical- Chemical Mechanisms of High-burnup Spent Nuclear Fuel (SNF) Processes at Elevated Temperatures and Degradation Behavior in Geologic Repositories

The overarching goal of the combined computational and experimental R&D activities proposed in this project is to enhance understanding of the mechanisms and thermal-mechanical-chemical (TMC) parameters controlling the instant release fraction (IRF) and matrix dissolution of high-burnup (HB; burnup) spent nuclear fuels (SNFs) and the subsequent formation, stability, and phase transformations of SNF alteration products under long-term storage and geological disposal conditions. Uranium dioxide may undergo oxidative corrosion/alteration, and the IRF may be increased for HB SNF, both of which may affect environmental systems associated with SNF long-term storage and disposal. The oxidative matrix dissolution may form various complex uranyl-based phases, including a rich variety of oxides, silicates, carbonates and other secondary minerals in varied geological environments (e.g., studtite, metastudtite, amorphous uranyl peroxide, uranium trioxide, triuranium octoxide, schoepite, dehydrated schoepite, metaschoepite, becquerelite, soddyite, rutherfordine,...). These uranyl phases generally have higher mobility UO 2 +2 species than less soluble U 4+ phases. However, limited information on the thermodynamic properties and formation kinetics of these uranyl-bearing phases is available to predict explicitly paragenesis under the conditions relevant to long-term storage or disposal. The proposed project draws on complementary expertise and research backgrounds from the team members: (i) to apply a combined ab initio modeling (UNLV/UTEP and SNL) and experimental (UNLV) strategy investigating the high-temperature TMC mechanisms of alteration of SNF under α-radiolysis conditions; (ii) to investigate the mechanistic of phase transformations in UNF degradation products under various conditions expected in long-term storage systems (e.g. (UO 2 )O 2 (H 2 O) 4 → (UO 2 )O 2 (H 2 O) 2 → U 2 O 7 → UO 3 → U 3 O 8 ); (iii) to determine high-accuracy TMC parameters for complex uranyl-based phases formed in storage or geological disposal environments (e.g. UO 3 (H 2 O) 2 , Ca[(UO 2 ) 6 O 4 (OH) 8 ] 8 H 2 O, (UO 2 ) 2 (SiO 4 ) 3 2H 2 O,…). The unforeseen COVID-19 pandemic led to the laboratory/campus closure since March 2020, that resulted in a significant delay in reaching milestones in a satisfactory manner, due to (i) the statewide recommendation from stop-working to later limited work in the lab and work-from-home (WFH), (ii) no in-person interactions, and (iii) a hiring freeze at UNLV. Therefore, a no cost extension (10/01/2021- 9/30/2022) was requested to help make up the time we lost during the global pandemic in 2020-2021, leading to paradigm shifts in the focus of the project in the following three main tasks: Task 1 (Computational), Task 2 (Experimental), and Task 3 (Final report, due on 12/29/2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing iron-chromium-aluminum (FeCrAl) alloys in place of zirconium-based alloys in major core structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing the chromium-coating on major zircaloy structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Design of a Loss of Coolant Blowdown Capsule for Remote Assembly with High Burnup Fuel

The Transient Water Irradiation System is an enhanced capability capsule type irradiation vehicle designed to support fuel safety research for light water reactor specimens in the Transient Reactor Test Facility and is designed to simulate loss of coolant and reactivity-initiated accidents. The capsule was designed, deployed, and commissioned with fresh fuel specimens to validate instrumentation and prepare for previously irradiated high burnup specimens. The irradiation system features an extensive in-situ instrumentation package to detect phenomena typical to light water reactor fuels. To accommodate the assembly with high burnup specimens inside the Hot Fuel Examination Facility, and to ensure instrumentation integrity is maintained throughout assembly, the design was updated to support remote handling. The updated design features a hinge mechanism which allows for remote pre-irradiated specimen loading and assembly while protecting sensitive instrumentation by relocating during loading. Fixtures and equipment have been developed to handle the experiment capsule and components in cell, and to support remote assembly. The experiment module also includes radiation shielding and contamination control to support operations after removal from the hot cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DEVELOPMENT OF TRANSIENT FISSION GAS RELEASE CAPABILITIES TO STUDY HIGH BURNUP COMMERCIAL FUEL PERFORMANCE UNDER LOSS OF COOLANT ACCIDENT CONDITIONS

The contribution of transient fission gas release to the burst behaviour of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant, additional transient fission gas release in excess of steady state fission gas release from normal operation. The timing of this release and the various conditions that impact this release warrant further study. To better characterize this phenomenon, a system was developed to capture and characterize fission gas released from irradiated fuel during simulated LOCAs. This system is an upgrade to the existing Oak Ridge National Laboratory Severe Accident Test Station LOCA test apparatus and detects fission gas by gamma counting cryogenically cooled cold traps and by using an online gas mass spectrometer. The examined fuel can be pressurized or open to the furnace atmosphere. The design and operation of this new capability are discussed. Results from the first tests on high burnup fuel are presented. In the initial test, the LOCA relevant transient fission gas release was 5.3% for an unpressurized segment. Post-test microscopy on the tested samples is presented to demonstrate the likely radial source of transient fission gas release.

Harp, Jason↗

Loss of Coolant High Burnup (LOC-HBu) Program

LOC-HBu is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments II (FIDES-II). LOC-HBu is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under loss of coolant accident (LOCA) conditions. LOC-HBu aims to support burnup extension needs by addressing identified R&D priorities to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. The need for such experiments was identified in the United States by the Fuel Performance Testing Technical Expert Group (FPTTEG) under the Electric Power Research Institute (EPRI)’s Collaborative Research on Advanced Fuel Technologies (CRAFT) framework. A combined in-pile and furnace testing program plan was developed by researchers at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) which was reviewed and approved by the FPTTEG Combined TREAT-LOC and SATS LOCA Experiment Plan. The first 4 tests identified in this plan are being put forward by the U.S. Department of Energy (U.S. DOE) and INL for inclusion in the FIDES-II second triannual work period. The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program was specifically developed to address data gaps and opportunities identified via detailed review of the existing public knowledge base on LOCA FFRD, as well as reviewing specific experimental development activities regarding prototypic LOCA conditions for light-water reactor (LWR) systems. The program will systematically investigate the impacts of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions not yet fully investigated. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Numerical Method Improvements in Griffin for Pebble Bed Reactors with a Focus on the Computation of Burnup

Griffin, a MOOSE (Multiphysics Object-Oriented Simulation Environment) based application targeting transient multiphysics modeling of advanced reactors, has been used recently to model both high-temperature gas-cooled and fluoride-salt-cooled pebble bed reactors (PBRs). Griffin uses deterministic methods for solving neutron transport and an Eulerian approach to model pebble movement. An Eulerian approach is also used to directly compute burnup instead of using a pass approach like other tools such as VSOP or PANGU. This work discusses verification efforts and numerical method improvements related specifically to the Eulerian modeling approach implemented for directly computing pebble burnup.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Advanced Electron Microscopy Characterization of the Fuel-Cladding Chemical Interaction Region in a High Burnup U-10Zr Fuel

To support the development of U-10Zr metallic fuel, advanced characterization techniques have been applied to the fuel-cladding chemical interaction (FCCI) region in a Na-bonded solid U-10Zr fuel cross-section that was irradiated to a burnup of ~13.1 at.% at the Fast Flux Test Facility. 17 phases were identified in the FCCI region through a combination of high-resolution scanning transmission electron microscopy (STEM), STEM based energy dispersive X-ray spectroscopy (STEM-EDS), and TEM based selective area electron diffraction (TEM-SAED) analysis. In this talk, we will also discuss the implications of results on metallic fuel FCCI by focusing on the formation of Zr rind, fission product migration, and HT-9 cladding integrality under the investigated thermal irradiation conditions. This work complements our previous study on the TEM characterization of the fuel region of this high burnup fuel sample and serves as scientific basis to support metallic fuel development and qualification.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A realistic TRACE PWR LOCA model with application to high-burnup analysis

This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (>62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. In conclusion, the model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding

Here, structure of the pellet/cladding-interaction (PCI) layer of a high-burnup nuclear fuel with niobium-bearing cladding was investigated using modern transmission electron microscopy (TEM) techniques. Morphology of the PCI consisted of uniform zirconia layer on the cladding side and finger-like features towards to the ceramic fuel. The PCI layer showed a complex microstructure that constitutes three distinct zirconia layers formed by two zirconia phases. From cladding to fuel, monoclinic, tetragonal, and another tetragonal phase were determined via TEM diffraction patterns and precession electron diffraction (PED) using ASTAR microscopy platform (AMP). Characterizations showed that monoclinic and tetragonal phases close to the cladding were crack-free, but the tetragonal layer adjacent to the fuel contained cracks and pores. Presence of the monoclinic layer suggested that its formation involved the gaseous diffusion of oxygen below threshold dose of fission product damage before the fuel/cladding contact occurred. Furthermore, high-resolution TEM revealed compound phase of (U,Zr)O 2 at fringes of zirconia fingers on the fuel side.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

Here, this study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A phase-field model of quasi-brittle fracture for pressurized cracks: Application to UO 2 high-burnup microstructure fragmentation

Here, in this paper, we present a phase-field model of quasi-brittle fracture with pressurized cracks, with dedicated applications for polycrystalline materials. The model is formulated as a minimization problem within the variational framework. The external work done by pressure on the crack surfaces is included in the objective function. Several careful modeling choices lead to a regularization-length-independent critical strength. The model is constructed to give a softening response with an underlying linear traction-separation law. The pressure-dependent softening response and the regularization of the prescribed pressure are demonstrated with a (quasi) one-dimensional numerical analysis. In a two-dimensional numerical analysis under plane strain assumptions, the critical stress corresponding to crack propagation (as predicted by our quasi-brittle fracture model) is compared with linear elastic fracture mechanics (LEFM) analytical solutions. Our model is further utilized to simulate fission-gas-induced fragmentation of the UO 2 high-burnup structure (HBS). Simulation results show that pressurized bubbles can cause crack nucleation and propagation, and that the bubble size and the surrounding external pressure affect the critical pressure corresponding to crack nucleation. Simulations of a partial HBS at different recrystallization stages show that different grain structures (due to recrystallization) also influence crack paths and fragmentation morphology.

36 MATERIALS SCIENCE↗

TRISO Burnup-Dependent Failure Analysis of a HTGR Design-Basis Accident Using BISON

Here, this work assesses the failure behavior of the silicon carbide (SiC) layer in TRIstructural ISOtropic (TRISO) fuel particles with BISON during both steady-state and transient conditions for the MHTGR-350 design by General Atomics. A one-dimensional BISON model for a uranium oxycarbide–bearing TRISO is developed to simulate a power level of 50 mW/particles up to 12.4% fissions per initial metal atom (FIMA). Stress predictions for the materials of interest are presented as a function of burnup, along with the SiC failure probability computed using Weibull statistics under the assumption of realistic SiC quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗