Search NASA⌕ Search

SEARCH · Search NASA

Results for “nuclear cladding”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,459 records · Page 81
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

COUPLING OF RELAP5-SCDAP MOD4.0 AND NEUTRONIC CODES

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Presentation of FP7 matter project: general overview

The 2010-2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), prepared in the frame of the Sustainable Nuclear Energy Technology Platform (SNETP), establishes a very tight time schedule for the start of construction of the European Gen IV prototypes; namely the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA will start in 2014 and that of the SFR Prototype ASTRID will start in 2017. The GEN IV reactors pose new challenges to the designers and scientists in terms of higher operating temperature, higher fuel burn-up, and in some cases more corrosive environment with respect to the present technologies and which impacts the materials performance. In this frame, the MATTER (Materials Testing and Rules) Project starts well targeted R and D activities to perform careful materials studies in GEN IV operational conditions and to find out criteria for the correct use of these materials in relevant reactor applications. Aim of the MATTER Project (that involved 27 partners and will end in 2015) is to complement the materials researches, in the frame of the European Energy Research Alliance (EERA) guidelines, with the implementation of pre-normative rules. The MATTER Project is divided in 3 technical Domains (called DM): DM1 - Development of test and evaluation guidelines for structural materials: to develop/establish best practice guidelines for testing and evaluation procedures, which are aimed to screen and characterize nuclear materials for innovative nuclear systems. DM2 - Pre-normative R and D for Codes and Standards: Pre-normative activities are performed, comprehensive of experiments, to revise and update the design rules (with an EU level consensus) in order to answer to some short term needs of the two projects ASTRID and MYRRHA with respect to the design and the construction of structural components. DM3 - Joint Program Scheme, implementation and Priorities: to optimise the effectiveness and efficiency of the EERA Joint Program on nuclear materials for innovative reactors and to support specific research activities related to fundamental understanding of ODS steels fabrication. ODS steels are considered candidate materials, in the medium-long term, for high fuel burn-up cladding application. After a brief presentation of DM1 and DM3, this paper mainly focuses on description of Pre-normative R and D activities for Codes and Standards (DM2). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

MO-FG-303-05: A Feasibility Study of Using a Cherenkov Detector Material with the Prompt Gamma Range Verification Technique in Proton Therapy

Purpose: To simulate the feasibility of a Cherenkov glass material for the determination of the penetration depth of therapeutic proton beams in water. Methods: Proton pencil beams of various energies incident onto a water phantom with dimensions of 5 x 5 x 30 cm{sup 3} were used for simulation with the Geant4 toolkit. The model used standard electromagnetic packages, packages based on binary-cascade nuclear model, several decay modules (G4Decay, G4DecayPhysics, and G4RadioactiveDecayPhysics), and optical photon components (G4OpticalPhysics). A Cherenkov glass material was modeled as the detector medium (7.2 g of In2O3 + 90 g cladding, density of 2.82 g/cm{sup 3}, Zeff = 33.7, index of refraction n(600 nm) = 1.56, and energy threshold of production Eth = 156 keV ). The emitted secondary particles are analyzed characterizing their timing, energy, and angular distributions. A feasibility analysis was conducted for a simplistic detector system using this material to locate the position of the Bragg Peak. Results: The escaping neutrons have energies ranging from thermal to the incident proton energy and the escaping photons have energies >10 MeV. Photon peaks between 4 and 6 MeV were attributed to originate from direct proton interactions with {sup 12}C (∼ 4.4 MeV) and {sup 16}O (∼ 6 MeV), respectively. The escaping photons are emitted isotropically, while low (≤10 MeV) and high (>10 MeV) neutrons are isotropic and forward-directional, respectively. The emissions of photons are categorized into prompt (∼ns) and delayed (∼min) where the prompt photons include the 4.4 and 6 MeV. The Cherenkov material had on average <2% of neutron interactions while LYSO and BGO scintillators had a minimum of ∼50%. Our simplistic detector system was capable of discerning Bragg Peak locations using a timing discrimination of ∼50 ns. Conclusion: We investigate the viability of using the Cherenkov material for MeV photon detection medium for the prompt gamma technique.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Simulation methodology for fuel assembly drop accident during handling

Fresh and irradiated fuel assembly handling is a routine part of Nuclear Power Plant activities. Fuel Assembly handling occurs during fresh fuel receipt, core refueling, spent fuel assemblies transfer from spent fuel pool to the storage containers, etc. Westinghouse, ENUSA and KEPCO-NF have collaborated in the technical area of Fuel Assembly Drop Accident (FADA) Simulation to update the approach for estimating the number of failed fuel rods as a consequence of a fresh or irradiated fuel assembly drop, which might occur during a handling accident in a PWR-type reactor. The proposed FADA simulation methodology starts with an assessment of fuel assembly impact velocity at the final stage of the drop in water or air. The vertical, horizontal or inclined drop directions are considered. This velocity is an input to determine a fuel assembly structural response during impact which defines the loading function for the fuel assembly components including spacer grids. The dynamic load on the fuel rod is calculated from the detailed spacer grid model and the corresponding loading function. Finally, the load from the spacer grid spring and dimples are applied to the fuel rod model to determine the cladding parameters for use in compliance to fuel rod cladding failure criteria. A combination of first principal, finite element analysis (FEA) and computational fluid dynamic (CFD) methods has been utilized to create a technically sound methodology. A number of specific and traditional tests for the fuel assembly and its components has been used to confirm that the proposed approaches which describe the physical phenomena related to the fuel assembly drop accident are reasonably conservative.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Oxide Morphology of a FeCrAl Alloy, Kanthal APMT, following Extended Aging at 300-600C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on a commercial FeCrAl alloy, Fe-21wt.%Cr-5wt.%Al-3wt.%Mo (Kanthal APMT). Aging treatments were performed for 100-1000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, isolated crystalline regions start to appear during short period aging time and expand with extended exposures. Crystalline α-Al2O3 oxide film develops at 600 °C and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

High temperature oxidation resistant iron-chromium-aluminum (FeCrAl) alloys are candidate alloys for nuclear applications due to their exceptional performance during off-normal conditions such as a loss-of-coolant accident (LOCA) compared to currently deployed zirconium-based claddings [1]. A series of studies have been completed to determine the weldability of the FeCrAl alloy class and investigate the weldment performance in the as-received (non-irradiated) state [2,3]. These initial studies have shown the general effects of composition and microstructure on the weldability of FeCrAl alloys. Given this, limited details on the radiation tolerance of FeCrAl alloys and their weldments exist. Here, the highest priority candidate FeCrAl alloys and their weldments have been investigated after irradiation to enable a better understanding of FeCrAl alloy weldment performance within a high-intensity neutron field. The alloys examined include C35M (Fe-13%Cr-5% Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions. Two different sub-sized tensile geometries, SS-J type and SS-2E (or SS-mini), were neutron irradiated in the High Flux Isotope Reactor to 1.8-1.9 displacements per atom (dpa) in the temperature range of 195°C to 559°C. Post irradiation examination of the candidate alloys was completed and included uniaxial tensile tests coupled with digital image correlation (DIC), scanning electron microscopy-electron back scattered diffraction analysis (SEM-EBSD), and SEM-based fractography. In addition to weldment testing, non-welded parent material was examined as a direct comparison between welded and non-welded specimen performance. Both welded and non-welded specimens showed a high degree of radiation-induced hardening near irradiation temperatures of 200°C, moderate radiation-induced hardening near temperatures of 360°C, and almost no radiation-induced hardening at elevated temperatures near 550°C. Additionally, low-temperature irradiations showed the non-welded specimens to exhibit strain-induced softening (decrease in the true stress level) with increasing plastic strain during tensile testing. Fracture for the weldments was found to occur exclusively within the fusion zone. The mechanical performance of the weldment was speculated to be directly linked to variances in the radiation-induced microstructure including the formation of dislocation loops and precipitation of the Cr-rich α' phase. The localized microstructural variation within the weldments, including grain size, was determined to play a significant role in the radiation-induced microstructure. The results summarized within highlight the need for additional data on the radiation tolerance of weldments as the mechanical performance of the fusion zone was shown to be the limiting factor in the overall performance of the weldments.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Oxide Morphology of C26M at 300 - 600 °C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on an Oak Ridge National Laboratory developed Fe-12Cr-6Al- 2Mo-0.2Si-0.03Y alloy (designated ‘C26M’). Aging treatments were performed for 100-2000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, crystalline α-Al 2 O 3 oxide film develops and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening on the cladding inner surface during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Overview of a Versatile Loading System for Anisotropic Material Property Characterization

Additive manufacturing, cold rolling, and other thermomechanical treatments on metals, especially HCP, can cause texture and oriented grain structures resulting in anisotropic mechanical properties. This can lead to macroscopic response that significantly deviates from isotropic assumptions which motivates studying mechanical properties in multiple directions. In addition, multiaxial loading and complex stress states overlapping the above mentioned material anisotropy can lead to unexpected outcomes. Idaho National Laboratory developed an advanced mechanical testing system to study a range of uniaxial to multiaxial stress states while measuring the anisotropic response bringing insight into the overlap between material properties and stress states. High-temperature capability and stress- or strain-controlled loading is available to enable a variety of experiment types and conditions to measure elastic, plastic, and viscoplastic properties. This poster presents the design and capabilities of this system with preliminary results highlighting the benefits of multiaxial loading and anisotropic analysis in an integrated system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AFC Metal Fuel Workshop Presentations

Compilation of PowerPoint slide decks from presentations given at the 3rd Annual AFC Metal Fuels Workshop. The purpose of this workshop is to bring the national metallic fuel stakeholders together to discuss status and needs of metallic fuel and review the DOE AFC Metal Fuel R&D. Desired Outcomes - Advance Metallic Fuel Technology: 1. Stakeholder Community Building 2. National Program R&D Accountability, Planning Support, and Increased Stakeholder Advocation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of molecular dynamics potential for uranium silicide fuels

Use of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. For example, silicide compounds (U3Si2 and U3Si) are known to become amorphous under irradiation. There was clear independent experimental evidence to support a crystalline to amorphous transformation in those compounds. However, it is still not well understood how the amorphous transformation will affect on fuel behavior. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In order to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters and increase the predictive capabilities of the developed model, inputs from atomistic simulations are essential. First-principles based density functional theory (DFT) calculations will provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we will present our recent progress in developing a U-Si potential from ab initio data. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molybdenum Sleeve Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.55 cm Pitch)

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM-096, LEUCOMP-THERM-097, LEU-COMP-THERM-101, and LEU-COMP-THERM-102. The purpose of these experiments was to measure the effects of molybdenum in nearly-critical systems. The molybdenum was introduced into the fuel arrays as tubular sleeves that surrounded some of the fuel rods in the fuel arrays measured. Four hundred molybdenum tubes nominally 12.7 mm outside diameter, 498 mm long, with 0.762 mm wall thickness were provided for the experiments by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). Small polyethylene adapters at each end of the tubes were used to center each tube on a fuel rod in the assembly. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 1261 fuel rod positions on a pitch of 0.610 in (1.5494 cm) in a series of 20 hexagonal rings surrounding the central fuel rod. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The five critical experiments in this series were performed in August through December 2022, in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility. Case 1 had no molybdenum sleeves, Case 2 had 208 molybdenum sleeves clustered at the center of the array, Case 3 had 397 molybdenum sleeves clustered at the center of the array, Case 4 had 175 molybdenum sleeves in the central position and in five alternating hexagonal rings, and Case 5 had 331 molybdenum sleeves in the central position and in seven alternating hexagonal rings. All five critical experiments are judged to be acceptable as benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of a U-19Pu-10Zr fuel performance benchmark case based on the IFR-1 experiment

Metallic nuclear fuels are subject to research and development for use in advanced reactors. Robust, accurate metallic fuel performance models are important for the design, safety analysis, and licensing of these reactors. However, metallic fuel performance models require additional development; they are not as mature as UO 2 fuel performance models. A benchmark case based on the IFR-1 experiment was developed to better gauge the accuracy of existing models, identify models for high-priority development, and potentially quantify any future improvements made by further model development.This work collected publicly available information on the IFR-1 experiment and used it to develop the benchmark case. Fuel behavior during the IFR-1 irradiation was simulated by using the fuel performance code BISON, and the predicted results were compared with postirradiation examination data from the IFR-1 experiment. Furthermore, a sensitivity study and tuning studies were performed as a preliminary investigation into the causes of inaccurate temperature and dimensional change predictions.The benchmark predicted reasonably accurate values for the burnup and fission gas release. There was error in the predicted temperatures, which could be explained by uncertainty in the input parameters and legacy temperatures. BISON over-predicted dimensional changes in the fuel and cladding. The sensitivity study showed that the dimensional changes were most sensitive to the fuel swelling anisotropy and the cladding void swelling model. Future benchmark and model development should focus on cladding swelling behaviors to improve dimensional change predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Accident Tolerant Fuel Analysis

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Final Seismic Shake Table Test Plan

The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test on the dry storage systems of spent nuclear fuel (SNF) to close the gap related to seismic loads on fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. Full-scale testing is needed because a dry storage system is a complex and highly nonlinear system making it hard to predict (model) the responses to seismic excitations. The non-linearity arises from the multiple spatial gaps in the system – between fuel rods and the basket, between the basket and dry storage canister, between the dry storage canister and the storage cask (overpack), and ventilation gaps. The non-linearities pose significant limitations on the value of tests with scaled systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗