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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Development of FeCrAl-ODS steels for ATF cladding

As an enhanced accident tolerant fuel (ATF) for LWR at severe accident condition, FeCrAl-ODS ferritic steels have been developed, and their applicability is being studied under the Japanese MEXT program. Alloying design of FeCrAl-ODS ferritic steels has been conducted, focusing on effect of mutual interaction of Cr-Al on oxidation resistance and α'-embrittlement. Effects of co-addition of Zr and excessive oxygen have also been studied to improve high-temperature strength. The ring tensile test of the manufactured claddings showed that tensile strength increases by Zr-addition and Zr-added FeCrAl-ODS steels have the highest strength among the existed steels at entire temperature range from room temperature up to 1000 Celsius degrees. Although Zr-addition sometimes leads to growth of alumina scale by rapid inward diffusion of oxygen through zirconia inside alumina, oxidation rate of Zr-added FeCrAl-ODS steels was successfully reduced to the level of Zr-free specimen by controlling oxygen content to form Y{sub 2}Zr{sub 2}O{sub 7} inside ODS steel matrix. This finding makes possible to improve not only high-temperature strength but also steam oxidation resistance in the Zr-added FeCrAl-ODS steels. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2019

Systematic Technology Evaluation Program for SiC-based BWR Channel Box

Silicon carbide (SiC) –based accident-tolerant fuel (ATF) cladding of light water reactors (LWRs) has become a topic of great interest in nuclear material development since the Fukushima Dai-ichi Nuclear Power Plant accident. Extensive development of SiC cladding has identified benefits of the concept, including its inherent accident resistance and neutronic performance. Advancements achieved during the development of SiC-based ATF cladding have made it attractive for use in accident-tolerant cores of LWRs. In this document, we consider the use of SiC-based materials to replace the zirconium alloy used for the channel boxes of boiling water reactors. Both accident resistance and prevention of the bowing observed in current grades of channel boxes are needs driving the exploration of a new channel box system. To provide a pathway for channel box development, this document lays out (1) a current technology overview, (2) the technology gap, and (3) a systematic technology evaluation plan. The objectives of the technology evaluation plan are to address key feasibility issues revealed by the knowledge gap analysis and to contribute to advancing the technology readiness level. Further development will be required in the technology implementation phase. The technology evaluations include understanding the behavior of channel boxes under both normal and off-normal conditions, identifying potential mitigation strategies for the issues if necessary, and improving channel box performance based on the knowledge obtained. The structure of the technology evaluation plan and each task are shown in Figure S.1. The knowledge gap analysis identified key research areas, including irradiation effects, hydrothermal corrosion, accident behavior, and design and manufacture. These are common issues between SiC-based channel boxes and SiC-based fuel cladding, although technical evaluations specific for the channel box components are required. This document also describes details of the technical feasibility evaluation methods and plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

FAST-1.2.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation (including creep), fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of Cone Wedge Ring Expansion Test to Evaluate Mechanical Properties of Clad Tubing Structure

To determine the hoop tensile properties of irradiated fuel cladding in a hot cell, a cone wedge ring expansion test method was developed. A four-piece wedge insert was designed with tapered angles matched to the cone shape of a loading piston. The ring specimen was expanded in the radial direction by the lateral expansion of the wedges under the downward movement of the piston. The advantages of the proposed method are that implementation of the test setup in a hot cell is simple and easy, and that it enables a direct strain measurement of the test specimen from the piston’s vertical displacement soon after the wedge-clad contact resistance is initiated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

Gentry, Cole A↗
DOE OSTI2017

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED URANIUM (LEU) FUEL

The United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) is working to convert research reactors globally from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its predecessor programs have converted or verified the shutdown of 88 HEU research reactors and isotope production facilities, using LEU fuel developed in the 1980s. However, there is a small set of high performance research reactors which require a new high density LEU fuel for conversion, including six U.S. high performance research reactors (USHPRRs), which include the Advanced Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor (MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor (MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA. These reactors will share the same uranium-molybdenum (U-Mo) fuel plate design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with aluminum cladding. Three different application methods for the Zr diffusion barrier layer are being pursued – co-rolled, electroplating and plasma spraying. Fabrication of the higher density LEU U-Mo fuel has presented several challenges. Because the U-Mo fuel fabrication process uses the existing stockpile of 93% highly enriched uranium (HEU) to be blended with molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of enriching to 19.75% will not be realized. Homogeneity of the U-235 and molybdenum in the metal foil is a challenge due to U-Mo casting process results in variations between adjacent cast U-Mo plates. LEU will not require less U-235 as previously believed due to self-shielding properties of U-238 requiring additional U-235 to be loaded in each fuel element. Although the overall dimensions of the ATR fuel element remain unchanged, the weight of the LEU ATR fuel element is 72% greater than a HEU ATR fuel element. During waste recovery and recycling efforts, separation of aluminum cladding from the HEU fuel meat is a relatively simple chemical dissolution, however, separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly more challenging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, such as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.

expansion plug test↗
DOE OSTI2018

Plastic flow and microstructural instabilities during high-pressure torsion of Cu/ZnO composites

Highlights: • Hybrid bulk metal - ceramic nanoparticle material was processed by HPT. • ZnO nanoparticles exhibited significant plasticity of up to 86%. • Hydrodynamic instability determined the flow mode of ZnO particle clusters. • Microstructural stability of Cu decreased due to the co-deformation with ZnO. - Abstract: We conducted high-pressure torsion (HPT) of hybrid materials composed of two copper disks with a layer of ZnO ceramic nanoparticles sandwiched in-between. A detailed microstructural analysis of the processed samples revealed a surprisingly high level of plastic deformation introduced in individual ZnO particles, even though they possess a much high hardness than the surrounding Cu. Moreover, we demonstrated that by tailoring the initial geometry of the system (i.e. introducing holes in the cladding Cu disks), we can effectively change the flow mode of ZnO particle clusters from laminar to turbulent one and facilitate their deagglomeration process. Finally, the presence of ZnO particles decreased the microstructural stability of the Cu matrix. We discussed the obtained results in terms of hydrodynamic and microstructural instabilities during plastic deformation. This work expands our understandings of the cooperative severe plastic deformation of two dissimilar materials and sheds new light on the design of hybrid materials.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2013

Status report of advanced cladding modeling work to assess cladding performance under accident conditions

Scoping simulations performed using a severe accident code can be applied to investigate the influence of advanced materials on beyond design basis accident progression and to identify any existing code limitations. In 2012 an effort was initiated to develop a numerical capability for understanding the potential safety advantages that might be realized during severe accident conditions by replacing Zircaloy components in light water reactors (LWRs) with silicon carbide (SiC) components. To this end, a version of the MELCOR code, under development at the Sandia National Laboratories in New Mexico (SNL/NM), was modified by replacing Zircaloy for SiC in the MELCOR reactor core oxidation and material properties routines. The modified version of MELCOR was benchmarked against available experimental data to ensure that present SiC oxidation theory in air and steam were correctly implemented in the code. Additional modifications have been implemented in the code in 2013 to improve the specificity in defining components fabricated from non-standard materials. An overview of these modifications and the status of their implementation are summarized below.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Design and analysis of oxidation tests to inform FeCrAl ATF severe accident models

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhan ced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as ch annel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl aloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Implementation of an Orificing Optimization Algorithm in the DASSH Subchannel Analysis Code

The Ducted Assembly Steady-State Heat transfer code (DASSH) performs full-core subchannel thermal hydraulics calculations in liquid metal fast reactors. One of the applications of subchannel codes is to optimize coolant flow orificing. As a design activity, the primary task is to determine the best way to divide assemblies into groups and distribute coolant flow rates among them. This report documents an algorithm implemented in DASSH to automatically optimize coolant orificing. Over the course of multiple iterations, DASSH determines the orifice grouping and flow distribution that minimizes peak coolant, clad, or fuel temperatures across all timesteps for a user-specified number of assembly groups. The total coolant flow rate in the reactor is constrained to achieve the specified core-average outlet temperature. The flow rate to each orifice group may also be constrained by the allowable pressure drop. The distribution of coolant flow among groups is accelerated using a predictor-corrector algorithm based on interpolated results from single-assembly parametric calculations. The assembly orificing grouping is initially predicted based on assembly power but can be refined if results demonstrate that an assembly would fit better in another group. The algorithm is demonstrated with two case studies. The first is a simple model for a reactor core consisting of just fuel assemblies; the pin power distributions are specified to create a situation where the initial assembly grouping prediction is suboptimal. This example is used to describe the initial grouping, demonstrate convergence over multiple iterations, and highlight the impact of regrouping. Then, the algorithm is applied to minimize peak clad and fuel temperatures in an example sodium-cooled fast reactor, the Versatile Test Reactor. The multicycle optimization confirms prior calculations for the reference core design. The example highlights how optimizing for different peak temperatures affects the results and demonstrates the use of the pressure drop constraint to limit the maximum flow rate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Polarized 3He Gas Circulating Technologies for Neutron Analyzers

We describe the development of an integrated system for quasi-continuous operation of a large volume neutron analyzer. The system consists of a non-magnetic diaphragm compressor, a prototype large volume helium polarizer, a surrogate neutron analyzer, a non-depolarizing gas storage reservoir, a non-ferrous valve manifold for handling gas distribution, a custom rubidium-vapor gas return purifier, and wire-wound transfer lines, all of which are immersed in a two-meter external magnetic field. Over the Phase II period we focused on three major tasks required for the successful deployment of these types of systems: 1) design and implementation of gas handling hardware, 2) automation for long-term operation, and 3) improvements in polarizer performance, specifically fabrication of aluminosilicate optical pumping cells. In this report we describe the design, implementation, and testing of the gas handling hardware. We describe improved polarizer performance resulting from improved cell materials and fabrication methods. These improvements yielded valved 8.5 liter cells with relaxation times greater than 12 hours. Pumping this cell with 1500W laser power with 1.25nm linewidth yielded peak polarizations of 60%, measured both inside and outside the polarizer. Fully narrowing this laser to 0.25nm, demonstrated separately on one stack of the four, would have allowed 70% polarization with this cell. We demonstrated the removal of 5 liters of polarized helium from the polarizer with no measured loss of polarization. We circulated the gas through a titanium-clad compressor with polarization loss below 3% per pass. We also prepared for the next phase of development by refining the design of the polarizer so that it can be engineer-certified for pressurized operation. The performance of our system far exceeds comparable efforts elsewhere.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Expert Meeting Report: Cladding Attachment Over Exterior Insulation

The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. The location of the insulation to the exterior of the structure has many direct benefits including better effective R-value from reduced thermal bridging, better condensation resistance, reduced thermal stress on thestructure, as well as other commonly associated improvements such as increased air tightness and improved water management (Hutcheon 1964, Lstiburek 2007). The intent of the meeting was to review the current state of industry knowledge regarding cladding attachment over exterior insulation with a specific focus on: 1. Gravity load resistance, 2. Wind load resistance. The presentations explorethese topics from an engineering design, laboratory testing, field monitoring, as well as practical construction perspective. By bringing various groups together (who have been conduction research or have experience in this area), a more holistic review of the design limits and current code language proposals can be completed and additional gaps identified. The results of which will help informdesign standards and criteria.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2019

Update on the Thermal Hydraulic Investigations of a Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a canister in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the present investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and add to the existing knowledge base. The objective of the HDCS investigation is to capture the dominant physics of a commercial dry storage system in a well-characterized test apparatus for any given set of operational parameters. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗