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DOE OSTI2023

Conceptual Design of Engineering-Scale Molten Salt Spill Tests to Support Molten Salt Reactor Licensing

Nuclear accident progression analysis codes for reactor licensing must be validated using experimental data that was generated at a large enough scale so that the results are representative of full-scale accidents. This report describes a conceptual design of engineering-scale molten salt spill tests that will provide the necessary datasets to support molten salt reactor (MSR) licensing. Design considerations, salt compositions, test variables, processes to be quantified, and suitable measurement techniques are presented and discussed. Aspects of the conceptual engineering-scale molten salt spill test design presented in this report were made using insights from recent laboratory-scale molten salt spill tests conducted at Argonne. Additional input from model developers, MSR vendors, and technical experts will benefit the design of the proposed engineering-scale molten salt spill tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

On-the-Fly Energy Condensation for Whole-Core Multiphysics Simulations

As the nuclear energy industry considers ways to achieve improved economics in the current fleet of light-water reactors (LWRs), one possible approach is to operate each cycle longer. This causes a greater portion of the fuel to be burned and reduces the frequency of outages, which ultimately reduces the cost to operate the reactor. However, this also leads to higher burnup fuels than have traditionally been allowed in these reactors. There are concerns about integrity of high-burnup (HBu) fuel, especially during accident conditions such as loss-of-coolant accidents (LOCAs), as shown by Capps et al. To investigate these concerns, advanced modeling and simulation capabilities are under development to determine the susceptibility of HBu fuel to fuel fragmentation, relocation, and dispersion (FFRD). Improvements have previously been made to fuel performance capabilities to model these phenomena more accurately; multiphysics simulations have also been conducted to determine the power and burnup histories of the HBu fuel, which are needed as inputs to the fuel performance calculations. Most recently, new statistical approaches have been developed to identify a subset of fuel rods that have greater FFRD susceptibility, reducing the total number of fuel performance simulations required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

SCALE Modeling of the Sodium Cooled Fast-Spectrum Advanced Burner Test Reactor

This report documents the modeling and simulation of a sodium-cooled fast reactor (SFR) as part of a U.S. Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with the Oak Ridge National Laboratory code SCALE and the Sandia National Laboratories (SNL) code MELCOR. Based on publicly available benchmark specifications, a fully heterogeneous 3D SCALE model of the 250 MWth Advanced Burner Test Reactor (ABTR) was developed to demonstrate SCALE’s capabilities for full-core reactivity analysis, fuel inventory prediction, and decay heat analysis of an SFR. The benchmark specifications contain modeling details for the ABTR core at the beginning of equilibrium cycle (BOEC) at operating conditions; they were derived from a 2006 preconceptual design report produced by Argonne National Laboratory. The ABTR was designed to demonstrate reactor-based transmutation of transuranics, that is, to “burn” transuranics recovered from light-water reactor (LWR) spent fuel. The ABTR’s fuel is designed to operate in 4 month cycles using uranium/transuranic (U/TRU) metallic fuel, with a TRU content of approximately 20%, at a conversion ratio of approximately 0.6. Various reactivity calculations were performed with SCALE for the ABTR and, where possible, compared with results available in the open literature. Additionally, SCALE was used to perform a full-core depletion calculation over the 4 month cycle to obtain the nuclide inventory at the end of equilibrium cycle (EOEC). These nuclide inventories, decay heat, power profiles, and reactivity feedback coefficients at EOEC represent the initial conditions for analyzing severe accident scenarios with MELCOR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Multi-Physics System-level Simulations of a Generic Pebble Bed High-Temperature Gas Cooled Reactor with Coupled SAM/Griffin Model

A SAM model for a 200 MWth generic pebble bed high-temperature gas-cooled reactor (PBHTGR) is presented in this work. As an extension to the modeling effort from FY-22 (Ooi et al. (2022)), an improved core channel approach is used to model the pebble bed of the reactor. The improvement greatly simplifies the model and reduces the run time of the model. Furthermore, the SAM model is coupled to the Griffin reactor physics model from Stewart et al. (2021) using the MOOSE MultiApp system where neutronics calculations are performed by the Griffin model and thermal hydraulics calculations by the SAM model. Favorable predictions are produced by the SAM/Griffin coupled model for steady-state normal operating condition. A load-following transient is also simulated with the coupled model where the neutronics and thermal hydraulics responses of the core are predicted correctly by the model. A protected Pressurized Loss of Forced Cooling (PLOFC) accident is simulated with the SAM model, in which the reactor power is determined using the decay heat curve, and therefore the Griffin model is not needed and thus decoupled. A simplified RCCS (Reactor Cavity Cooling System) loop is added to the model to remove decay heat from the core during the transient. Qualitatively, the prediction by the SAM model matches the expected behavior of a PB-HTGR during a PLOFC accident.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Thermal Modeling Aluminum-Clad Spent Fuel Dry Storage Configurations [Slides]

A presentation covering the work done by the C430 Department at Idaho National Lab on thermal concerns for extended dry storage of aluminum clad spent nuclear fuel (ASNF). Computational fluid dynamics software Star CCM+ was used to model hypothetical accident conditions applied to a dry storage configuration containing ASNF. The hypothetical accident conditions being examined were a fire in which the dry storage configuration is fully engulfed in flames and a situation in which the vent ports of the overpack are completely blocked.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

This milestone report summarizes recent work to investigate higher fidelity modeling and simulation practices for large-break loss-of-coolant accident (LBLOCA) analysis in high-burnup pressurized water reactor (PWR) cores. Because of current industry interest in extending fuel cycle lengths ranging from 18 to 24-months, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been investigating the susceptibility of high-burnup core designs to fuel fragmentation, release, and dispersal (FFRD) during accident conditions such as LBLOCA. This work has prioritized developing and demonstrating a methodology for calculating core-wide susceptibility to FFRD and addressing uncertainties identified in the analysis and the US Nuclear Regulatory Commission (NRC) Research Information Letter on FFRD. Part of this investigation seeks to quantify differences between system-level thermal hydraulic behavior and higher fidelity subchannel modeling methods to identify potential safety concerns or opportunities to minimize FFRD susceptibility. To this end, the NEAMS subchannel code, CTF, is being used, along with the NRC system analysis code TRACE, for analysis of LBLOCA in a core containing high-burnup fuel. This project includes two thrusts: (1) improving on the existing TRACE model for a 4-loop PWR for LBLOCA so that the model is higher fidelity and consistent with current USNRC best practices, and (2) using CTF to perform pin-resolved modeling of the core region of the reflood phase of a LBLOCA in a high-burnup PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Application of NEAMS Multiphysics Framework for Species Tracking in Molten Salt Reactors

This report from Idaho National Laboratory (INL) summarizes the key modeling and simulation activities conducted under the Department of Energy (DOE) Molten Salt Reactor (MSR) Campaign during the Fiscal Year 2023 (FY23). The focus of the work was to leverage state-of-the-art modeling capabilities from the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) codes to enable novel multiphysics and multiscale modeling and simulation of MSRs. Through collaboration with NEAMS code developers, advanced multiphysics analysis capabilities for MSR systems were demonstrated by coupling depletion, thermal-hydraulics, and thermochemistry into an innovative framework for chemical species transport in MSRs. As a result, the framework can track nuclides throughout their lifetimes in the core, from production (depletion) to advection throughout the salt volume (thermal-hydraulics) and off-gassing or precipitation outside of the salt (thermochemistry). This work supports the near-term deployment of MSRs by integrating the synergistic efforts between the DOE’s MSR Campaign and NEAMS program. The resulting framework will help better connect system design modelers with experimentalists to better understand and predict complex physical behaviors in MSRs. Researchers and MSR developers alike can now leverage these new modeling and simulation capabilities to perform novel analyses with applications including: • MSR dynamics during normal operational transients and accident scenarios • Off-gas system design and performance for fuel cycle and depletion analysis • Corrosion and active chemistry control for reactor component health and lifetime determination • Source term, decay heat and activity determination in accident scenarios • Special nuclear material accountancy and chemical forensic analysis for safeguards • Digital twin development of experiments and experimental reactor demonstrations • Measurement requirements for instrumentation and control design • Uncertainty and sensitivity analysis of missing data to inform future experimental data collection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Simulation and Analysis on Reactor Pressure Vessel (RPV) subjected to Pressurized Thermal Shock (PTS) under SBLOCA scenario by using Cardinal to support the fracture mechanics analyses

The structural components that comprise nuclear reactors and their supporting structures are subjected to harsh operating environments that can challenge their integrity, especially after exposure for extended durations or under accident condition. As one of the most significant components of a Reactor, the Reactor Pressure Vessel (RPV) is exposed to an aggressive environment during the operation time (e.g. more than 40 years). Ageing degradation mechanisms (e.g. thermos-fatigue) could grow initial defects up to a critical size, increasing the susceptibility to failure in the RPV. The conventional methods are mostly based on simple crack and structure geometries. Very limited studies consider the real conditions of the RPV subjected to a thermal shock due to a Loss of Coolant Accident (LOCA). During a LOCA event, the most severe conditions take place when the emergency core cooling (ECC) water is injected inside the cold legs filled initially with hotter water and/or steam. The rapid cooling of the down-comer and the internal RPV surface followed probably by re-pressurization of the RPV causes large temperature gradients and variation of pressure which induces thermal-mechanical stresses. In order to develop the model for integrity assessment of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS), a multi-physics simulation, which includes the thermo-hydraulic, thermo-mechanical and fracture mechanics analyses is necessary. The multi-physics simulations are performed using Cardinal, a wrapping of the GPU-oriented spectral element Computational Fluid Dynamics (CFD) code NekRS and other multi-physics sub-modules within the MOOSE framework. Cardinal now fully supports MOOSE stochastic perturbations of NekRS models with varying boundary conditions, initial conditions, material properties, and any other quantity which is defined by a kernel (such as coefficients in a momentum source model). The implementation is designed in a flexible manner so that scalar values are sent from MOOSE into a user scratch space in NekRS, which can then be applied for any purpose within the NekRS case files (both on the host and device). When modeling PTS, several factors can impact the results significantly. In this report, the impacts of the geometry of the model, Reynolds number and buoyancy effect are investigated. Two geometry, i.e., a simplified model and a realistic RPV model, with both laminar and turbulent flow condition are adopted for the PTS simulation with and without buoyancy effect. The purpose of the investigation is to understand the impact of these factors on the prediction of temperature history of RPV. The accurate prediction on the temperature evolution, which will be exported to Grizzly code for further analyses on the progression of aging mechanisms and their effects on the integrity of RPV structures, is very crucial. Based on the understanding of these factors, a more sophisticated model is built to analysis the PTS under SBLOCA scenario. A literature survey is conducted to pick the SBLOCA scenario for the multi-physics simulation. The analysis helps to explain the form and the transformation of the cold plum when the ECC is activated under SBLOCA. This model can be can be applied to study the PTS effect for different RPV configurations. The results can help to assess structural component degradation for advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

MARVEL Hazard Evaluation ECAR-6440

MARVEL Hazards Evaluation evaluated the impacts of MARVEL operations, hazards, and postulated accidents. The hazard evaluation of MARVEL events and associated operations was performed for selection and evaluation of safety classification of systems, structures, and components (SSCs) and SSC safety functions, and for selection of design basis accidents (DBAs) applicable to the MARVEL microreactor design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

MARVEL 90% Final Design Report

This document provides documentation of the Microreactor Applications Research Validation and Evaluation Project’s (MARVEL) 90% Final Design, as required by U.S. Department of Energy (DOE) Standard-1189, “Integration of Safety into the Design Process." Per DOE-STD-1189-2016, the 90% Final Design documentation focuses on design completion, at a level capable of supporting procurement, construction, testing, and operation. At this phase, the design organization finalizes the hazards and accident analyses, Fire Hazard Analysis (FHA), security vulnerability assessments, and other supporting analyses for design completion. The objective of this report is to provide a high-level summary of the design thus far and provide references including, but not limited to, the following design deliverables: • Complete final drawings, specifications and commercial grade dedications that may be released for bid and/or construction. • Clearly defined testing plans for the safety and functionality of all subsystems. • Quality Assurance Program for Design, Testing and Procurement. • Software Quality Assurance Plan. • Code of Record (COR), applicable design requirements including codes and standards. • Final design that meets all the requirements stipulated in the COR. • Final design review, consisting of final validation of comment resolution from previous reviews, and a review of any additional developments since the last review. • Updated Safety Design Strategy. • Hazard Analysis. • Fire Hazard Analysis. • Accident analysis. • Security vulnerability assessment. • Current and detailed cost estimate. • Current construction schedule, and • Risk & Opportunities Assessment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Non-LWR Fuel Cycle Scenarios for SCALE and MELCOR Modeling Capability Demonstration

To assess the modeling and simulation capabilities for radionuclide characterization, criticality, and shielding in the nuclear fuel cycle of non-light–water reactor (LWR) technologies, the US Nuclear Regulatory Commission (NRC) initiated a collaborative project between the NRC, Sandia National Laboratories (SNL), and Oak Ridge National Laboratory (ORNL) with the goal to demonstrate capabilities of MELCOR and SCALE to calculate accident scenarios in all stages of the nuclear fuel cycle for relevant non-LWRs. The first project task was to develop representative nuclear fuel cycles and identify potential hazards and accident scenarios in the individual fuel cycle stages based on publicly available information. Because the nuclear fuel cycle is not established for any non-LWR concept, many assumptions were made, and it is anticipated that the details of the fuel cycles will eventually look different.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Multiphysics Pebble-Bed Reactor Control Rod Withdrawal Study

This work studied the responses of both a generic gas- and a fluoride-cooled PBR concept---the gPBR-200 and gFHR, respectively---during reactivity insertion accidents. Both models rely on 2-D axisymmetric simulations to solve the neutron flux distribution, nuclide concentrations, and temperature across the core---in addition to numerous representative pebble and TRISO particle simulations for determining fuel and moderator temperatures. This not only allows for computing maximum temperatures in the core---thus enabling estimation of how near the fuel is to peak operational and safety limits---as prescribed by specified acceptable fuel design limits, which are determined in such a way that fuel is not damaged during operational or anticipated abnormal occurrences---but also predicting how much of the core exceeds a given temperature limit, as well as determining the local energy deposition rate. These models consider both control rod withdrawal and ejection events. The former introduces a great deal more reactivity, as all the control rods are withdrawn (as opposed to a single one in the latter case), though at a much slower pace. In addition, for the gPBR-200, two limiting cases were considered: one with the core starting under hot full-power conditions and one with it starting under cold zero-power conditions. While the amount of reactivity added in the latter case is much higher (due to the far lower temperatures and the lack of neutron poisons such as Xe-135), the margin to temperature limits is also much more significant. Overall, for the design considered, none of the accidents resulted in the maximum fuel temperature reaching values close to the TRISO limit. However, the methodology presented herein could be very relevant if some designs consider reduced margins (e.g., higher temperatures) to achieve enhanced economics. Further model improvement is needed to better capture control rod worth, both in terms of cusping effects (as the rods are slowly withdrawn) and differential worth, especially as the tips of the rods near the upper cavity.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Mechanical creep testing and characterization of Cr coated Zircaloy

Zirconium based cladding has been developed for use in light water reactor fuel applications due to the material’s desirable low neutron scattering cross section. To help improve corrosion resistance in accident conditions, such as a loss of coolant accident (LOCA), a chromium coating can be applied to the outer diameter of the Zircaloy tube. Cladding strength at reactor relevant temperatures is important for the structural performance. Stress relaxation creep testing was performed on Cr coated Zircaloy-4 to measure strain rate as a function of stress for 350 and 650°C. These were performed at both 5 ± 1 % and 7.5 ± 1 % strain. The lower temperature provides results near normal operating temperature, while the higher temperature provides results relevant to off-normal conditions where recrystallization will take place. Characterization was also performed to analyze the structure of the Cr coating and Zircaloy-4 as well as the interface between the two.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Scoping Thermal Response Calculations of RNS Waste During Transport to and Disposal at the WIPP

Sandia National Laboratories (SNL) was contracted by the United States Department of Energy Environmental Management (DOE-EM), Los Alamos Field Office to perform mechanical and thermal scoping calculations as part of a study seeking to understand the ignitability risk of the Remediated Nitrate Salts (RNS) waste drums during transportation from the Waste Control Specialists (WCS) facility to Waste Isolation Pilot Plant (WIPP) and permanent disposal of the waste at WIPP. The scoping thermal simulations described in this report pertain to thermal calculations performed with a packaging system consisting of one Standard Waste Box (SWB) loaded with drums placed inside a Standard Large Box 2 (SLB2). During transportation, the SLB2 is inside Transuranic Package Transporter Model III (TRUPACT-III), which provides the third layer of the packaging. Once at the WIPP, it is assumed the SLB2 is extracted from the TRUPACT-III and maintained above ground, and then subsequently placed underground for permanent disposal. In these proposed configurations, the space between the SLB2 and the SWB is always filled by a layer of insulation consisting of air-filled glass microbubbles except for the bottom which rests directly on the SLB2. The thermal scoping calculations described in this report specifically address whether the introduction of external heat inputs, combined with the contributions from the internally generated radiolytic decay heat and chemical reactions, lead to an unstable thermal state during the time of its movement and placement in the permanent disposal location. The external heat inputs are of two forms: 1) ambient thermal irradiation (e.g., solar and ambient storage/disposal temperatures) and 2) accident-induced fire. Three scoping calculation scenarios were derived as representative, conservative scenarios: 1A) TRUPACT-III transient transportation, 1B) SLB2 48-hour outdoor storage with solar radiation, and 2) fully-engulfing fire during SLB2 handling or emplacement following a steady-state analysis in a 38 °C environment. All the simulated scenarios are conservative relative to the operational conditions expected for handling the waste package during transportation and placement in the WIPP underground disposal unit. The predictions obtained from simulating the three exposure scenarios revealed that adding the SLB2 and the air-filled glass microbubbles to the transport and storage/disposal configurations provides additional thermal protection of the drums beyond what the SWB provides alone, both during long-term above ground insolation and underground during a fire accident. Under the current transportation/storage/disposal concepts, the degree of protection provided by the packaging concept is sufficient to prevent the waste from being ignitable. The simulation results demonstrate that there is adequate margin to safely transport and place the RNS waste from WCS to the WIPP under the current operational concept.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

SSR COST – COmposite Structural Technologies for advanced reactor deployment (Final Scientific/Technical Report)

Moltex is designing a Generation 4 or Advanced nuclear reactor, the molten salt fast spectrum SSR-W. The design intent is to reduce the cost and complexity of the nuclear systems by relying on inherent safety characteristics, instead of engineered systems. One such characteristic as a natural tendency for power to decrease as temperature increases. Another characteristic that improves natural safety is the fact that the working fluids (molten salts) are at atmospheric pressure at operating temperatures. This significantly reduces the need for pressure boundary components and systems. Moltex is highly motivated to develop simple and economic reactors to reduce the carbon intensity and improve energy security of global electricity production. The COST project investigated the opportunities for reducing construction cost of advanced reactors with the adoption of Rigid Elastic Composite Civil works (RECC). The construction design and methods used in currently operational light water power plants would generally be suboptimal for the SSR-W since there is no requirement for civil structures to withstand high pressure accidents such as hydrogen combustion or steam explosion. In parallel to downselecting the most promising RECC-type technology, accident scenarios representing the largest credible challenge to the SSR-W civil and construction design were modelled. The capability of an SSR-W reactor constructed with and without the selected construction technology was evaluated and the consequential licensing challenges identified. A construction schedule and cost estimate of the SSR-W with and without the selected RECC construction technology was developed. Once the above data was collected, then Moltex was able to review the safety performance, license-ability and cost/schedule benefit of applying the preferred construction technology. The CAPEX/NPV benefit was quantified, and the licensing risks and opportunities were documented. Finally, the CAPEX/NPV benefit of the advanced construction technology was modest due to the suitability of traditional stick-built construction for the SSR-W design. However, previously unidentified benefits of the advanced construction technology were revealed during the studies that could credibly de-risk certain aspects of the design to be licensed. Since licensing (duration and delays) are very significant in new build project costs, the cost benefits may be significant. Further studies on the risk reduction of the construction technology are foreseen.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

A Comparison of Adverse Meteorology Inputs for Emergency Management and Safety Basis Dispersion Modeling Applications at SRS

The MELCOR Accident Consequence Calculation System (MACCS), used for safety basis accident analyses at SRS, implements adjustment factors to consider the increased vertical turbulence associated with local surface roughness and the increased lateral turbulence associated with time-based plume meander. The two models used for SRS emergency planning radiological consequence assessment, i.e., Hotspot and Puff/Plume, currently do not have an option to input or apply diffusion adjustments for meander or surface roughness. Adverse meteorology conditions at SRS have been determined as E stability class and 1.3 meters per second wind speed for ground level releases. This is the input used for emergency planning models. Contrastingly safety basis analyses use a different input, representing adverse meteorology values as F stability class and wind speeds of 1.3 meters per second (for a ground level release). This results in an apparent discrepancy between the two approaches. In this report we document the differing modeling assumptions, discuss the logic behind each modeling approach used to represent local roughness and plume meander conditions, and numerically compare dispersion coefficients and concentration estimates obtained from each set of model inputs.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2022

Multi physics simulation on Reactor Pressure Vessel (RPV) subjected to Pressurized Thermal Shock (PTS) by using Cardinal

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

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↗