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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.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Rapid depletion analysis of flowing-pebble reactor systems at equilibrium using SCALE

Several high-temperature gas-cooled reactor concepts (and more recently, salt-cooled designs such as the fluoride salt-cooled high-temperature reactor) feature core designs employing continuously circulating fuel pebbles. These reactor designs permit both continuous online refueling of fuel elements as well as higher overall achievable discharge burnups. However, rapid calculation of time-dependent fuel isotopic inventories proves challenging for this class of dynamic systems with current analysis tools. While iterative approaches employing coupled neutron transport have been developed to solve this issue, rapid depletion analysis techniques are needed to calculate time-dependent inventories for individual pebbles and batches (and thus the construction of full- core inventory at equilibrium). We propose a depletion analysis strategy for this type of system for cores at equilibrium. Drawing upon previous neutronic analysis of the PBMR-400 equilibrium core, we demonstrate the viability of developing collapsed one-group cross section libraries suitable for performing rapid depletion analyses with SCALE. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Light Water Reactor Sustainability Program Modeling of Cu Precipitate Contributions to Reactor Pressure Vessel Steel Microstructure Evolution and Embrittlement

This report summarizes the lower-length-scale effort during FY 2017 in developing simulation capabilities for the mesoscale: models for microstructure evolution and plasticity in reactor pressure vessel steels. During operation, reactor pressure vessels undergo hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation initiate on the atomic scale and manifest their eventual effects as degradation in the engineering scale properties. To predict the engineering scale property degradation, multiscale modeling and simulation capabilities are needed to better understand the microstructure evolution and to connect the microstructure feature evolution to the engineering scale material properties. In this report the development of coupling across different length scale models to understand the multi-physics property dedgration problem is discussed. The development of coupling a lattice kinetic Monte Carlo code to cluster dynamics to capture irradiation enhanced diffusion in the formation of defects is summarized. The implementation of a cluster dynamics model with Grizzly to predict Cu percipitate formation is discussed. A summary of the coupling of crystal plasticity and cluster dynamics to predict the effect on reactor pressure steel hardening is given.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Reactor System Safety: A case for New and Advanced Reactors

This presentation covers a generic overview of reactor system safety testing, analysis, and related research and development (R&D) focusing on new and advanced reactor systems. This presentation is prepared for graduate-level student seminar talks. No specific reactor design information is provided. Only publicly available information and related published articles and book contents are utilized.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2019

Multiscale model based design of an energy–intensified novel adsorptive reactor process for the water gas shift reaction

In this work, an adsorptive reactor (AR) process is considered that can energetically intensify the water gas shift reaction (WGSR). To best understand AR process behavior, a multiscale, dynamic, process model is developed. This multiscale model enables the quantification of catalyst and adsorbent effectiveness factors within the reactor environment, obliviating the commonly employed assumption that these factors are constant. Simulations of the AR's alternating adsorption-reaction/desorption operation, using the proposed model, illustrate rapid convergence to a long-term periodic solution. The obtained simulation results quantify the influence of key operating conditions and design parameters (e.g., reactor temperature/pressure, W cat /F CO , W ad /F CO , F H2O /F CO ratios, and pellet size) on the AR's behavior. Finally, they also demonstrate, for pellet diameters used at the industrial scale, significant temporal and axial variation of the catalyst/adsorbent pellet effectiveness factors. Finally, the energetic intensification benefits of the proposed AR process over conventional WGSR packed-bed reactors are quantified.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Enhancing the Chemical Energy Flux in a High-Temperature Tubular Counterflow Solid Fuel Synthesis Reactor Using a Bypass

Redox reactions of metal oxides offer a path towards using intermittent renewable resources for high-density thermochemical energy storage. Thermochemical energy storage often involves the flow of a particulate media. We describe a novel method to increase the throughput in a gravity-driven high-temperature thermochemical storage reactor flowing pelletized MgMnO. The moving bed reactor operates under counter-flow conditions and encounters particle flowability problems at temperatures of 1500 °C leading to sintering of the bed. Inertial forces of a counter-flowing gas can overcome the gravitational forces on the particles and limit the chemical energy storage rate of the reactor. We found that the insertion of a gas bypass (a slotted tube) into the reactor results in a 100% increase of the flow rates and achieved a 50% higher chemical energy storage flux compared to the operation without a bypass tube while mitigating the effects of sintering on the particles. As a result, the higher solid flow rates require a longer heated zone to reach a comparable residence time and extent of reduction compared to the lower flow rates.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Design Studies for the Optimization of 238 Pu Production in NpO 2 Targets Irradiated at the High Flux Isotope Reactor

Efforts to reestablish a domestic 238 Pu production capability in support of National Aeronautics and Space Administration mission objectives are ongoing throughout the U.S. Department of Energy complex. The Plutonium-238 Supply Project (PSP) was initiated in response to a report published by the National Research Council in 2011 stating that “without a restart of 238 Pu production, it will be impossible for the United States, or any other country, to conduct certain important types of planetary missions after this decade.” The PSP is targeting a sustained, constant production rate of 1.5 kg/year of heat source PuO 2 for several years. Design and optimization studies of 237 Np-bearing targets are underway at Oak Ridge National Laboratory (ORNL). It is anticipated that targets will be irradiated in ORNL’s High Flux Isotope Reactor (HFIR) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory. A variety of target materials, containments, arrangements, and irradiation histories have been analyzed, and the results indicate that a sufficient quantity of 238 Pu can be produced in HFIR and ATR to fulfill the PSP’s constant production rate target. This paper focuses on the design and optimization of new target configurations containing pellets that are (1) ~93% of the theoretical density of NpO 2 , (2) loaded into pins of cladding materials that can be handled as solid waste following postirradiation 238 Pu recovery operations, (3) irradiated in various vertical experiment facility (VXF) locations in the HFIR permanent beryllium reflector, and (4) rotated within and/or moved to another VXF location following each HFIR operational cycle to maximize 238Pu production and minimize peak heat generation rates.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Experimental and modeling study of chemical-based strategies for mitigating dust formation in fusion reactors [Plasma Flow Reactor Chemical Kinetics: A Method Toward Optimizing Nano-Manufacturing and Laser Inertial Fusion Energy Strategies]

We studied carbon/hydrogen/oxygen chemical kinetics at time scales and thermal conditions relevant to fusion energy applications using a custom-built plasma flow reactor to investigate chemical-based strategies for eliminating carbon dust formation in fusion reactors. Acetylene and oxygen gases under varying conditions of initial concentrations are injected into an inductively coupled argon plasma where complete molecular dissociation occurs. The evolution of chemical species is investigated along the plasma flow reactor as a function of temperature and residence time. Atomized species of C, H, and O cool from 5000 K to 1000 K within 30 ms at atmospheric pressures. We employed optical emission and infrared absorption spectroscopy to measure the reaction intermediates (e.g. C 2 ) and products (e.g. C 2 H 2 ). Chemical equilibrium models are inadequate to describe the evolution of carbon molecular products, and thus a chemical kinetics model is developed. In both experiments and kinetic modelling, we find that the addition of oxygen in 1:1 proportion to carbon strongly favors the formation of CO, preventing the formation of acetylene (an important soot precursor) in less than 10 milliseconds. The kinetics model is also used to perform a reaction sensitivity and a rate of production analyses to identify the rate determining steps and the major chemical pathways that control the acetylene production/consumption. The results demonstrate the feasibility of chemical-based strategies for eliminating the formation of carbonaceous particles in fusion energy reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Light Water Reactor Sustainability Program: Technical and Economic Considerations for Uprate of Existing Nuclear Reactors with Cogeneration

The United States nuclear reactor fleet consists of 63 pressurized water reactors and 31 boiling water reactors and is a pivotal component in the nation's energy infrastructure, supplying approximately 97 GW e of clean power. With the country's commitment to decarbonization by 2050, these reactors are not only instrumental in decarbonizing the electricity grid but also play a critical role in decarbonizing industrial processes, producing clean fuels, and scaling up CO 2 removal. This report delves into the potential for power uprates in the existing fleet to contribute to these decarbonization efforts, focusing on the expansion of capacity for applications such as hydrogen production and carbon capture and sequestration. Building on previous research, the report explores regional market demands for hydrogen, oxygen, and carbon dioxide, financial implications of oxygen and CO 2 sales from high-temperature steam electrolysis systems, and the potential for direct air capture systems paired with uprated nuclear plants.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
NASA NTRS2005

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .<el, so fluid movement and temperature changes will cause very minor effects). In previous SAFE-100 tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Bragg-Sitton, Shannon M.↗
DOE OSTI2022

Demonstration of MOOSE-based Griffin reactor physics, code for heterogeneous lead-cooled fast reactor analysis

The MOOSE-based reactor physics code Griffin was assessed on a heterogeneous pin-resolved model of a prototype lead-cooled fast reactor assembly. This model was developed in preparation for future use in MOOSE-based multiphysics calculations for computing hot channel factors. Heterogeneous multigroup cross sections were prepared using the fast reactor multi-group cross section processing code MC{sup 2}-3 using a two-step method. Griffin simulations were performed using the DFEM-SN solver on 576 cores on Argonne's LCRC cluster. Diffusion-based acceleration methods were applied (NDA and CMFD). Reference solutions were generated with continuous energy MCNP and the hybrid MOC/finite element solver PROTEUS-MOC for code-to-code comparison. Space-angle convergence studies were conducted to observe convergence in k-eigenvalue and axial pin power distributions. The fully resolved Griffin calculation was within 68 pcm of the MCNP eigenvalue and exhibited max 1.2% relative error in the axial pin power distribution. Griffin produced nearly identical results to PROTEUS-MOC when using the same 9-group multigroup cross-section set. Griffin demonstrated favorable scaling in wall-clock time and memory usage when using diffusion-based acceleration methods. Griffin is capable of simulating the pin-resolved heterogeneous LFR assembly with good accuracy and performance, and is suitable for future use in coupled high-fidelity hot channel factor simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiscale thermal hydraulic coupling methods for boiling water reactor simulation

During the last two years, the Virtual Environment for Reactor Applications (VERA) has been extended to simulate boiling water reactors (BWRs). The thermal hydraulic effects present in BWRs are far more complex than those in pressurized water reactors (PWRs). Therefore, the runtime is significantly increase and convergence behavior worse compared to PWRs. Most of the additional runtime is spent during the first few coupled iterations, when the power shape is still rapidly evolving, dramatically affecting the thermal hydraulics (TH). To alleviate the increased computational expense, a multiscale TH coupling approach was developed in VERA in which a highly efficient, simplified TH model is solved for several coupled iterations until the power shape is partially converged. The simplified solution is used to inform the assembly-wise flow distribution in the high-fidelity TH solver CTF, reducing the amount of work required to properly balance the pressure drop in each channel and reducing runtime for the couple calculation. Because the final TH calculations are performed with CTF, there is ultimately no impact on the accuracy of the converged solution. This paper presents the details of this multiscale TH coupling approach, along with results for a 4x4 array of GE-14 fuel bundles and whole-core coupled simulations of the Hatch core. These two cases show that the multiscale approach dramatically reduces the runtime of coupled BWR simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

The international research reactor conversion effort and its contribution to the validation of reactor physics and thermal-hydraulics codes

Full text of publication follows. With about 200 reactors in operation, civilian Research and Test Reactors (RTRs) represent approximately a third of the global nuclear fleet. RTRs make use of core designs that are drastically different from commercial power plants to perform a wide variety of non-power applications that greatly benefit society. Because they often rely on high neutron fluxes, RTRs are designed with relatively compact cores and as a result, prior to the 1980's, were often deployed using Highly Enriched Uranium fuel (HEU, {sup 235}U/U = 20 wt. %). Due to proliferation concerns, the international community aims at eliminating the use of HEU in civilian facilities and favor instead the use of Low Enriched Uranium fuel (LEU, {sup 235}U/U < 20 wt. %). A program to support conversion of the world's RTRs to LEU fuel has been initiated in 1978 by the U.S. Department of Energy (DOE). This program is still alive today and has achieved more than 103 conversion metrics. Today, the program focuses heavily on the conversion of so-called high-performance RTRs, which are far more challenging than previous conversions as they require new fuel element designs and the use of new, higher density LEU fuel forms. Development and qualification of new LEU fuel elements is ongoing and requires extensive engineering analysis and testing. Both activities require the development and validation of codes and methods for reactor physics, thermal-hydraulics, and multi-physics, which in turn rely on experiments performed in RTRs or other experimental facilities. This talk will present the collection of RTR experimental data and benchmark analyses from the international conversion program that contribute to the validation of computer codes and methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

01-07 Reactor Supplemental Shielding For the NRIC DOME Test Bed 01-07 Reactor Supplemental Shielding For the NRIC DOME Test Bed 2

A radiation shielding system is needed to protect personnel and equipment from harsh neutron and photon fluxes during: Reactor operations Post shut down operations Disassembly Decommissioning Examples of RSS driving requirements Radiological Controls Operating dose rate: the dose rate during normal reactor operations shall be <0.5 mrem/hr at the exterior wall of DOME containment Outdoor radiation limit: shielding shall provide the dose to a facility worker outside of DOME to <0.05 mrem/hr at 50 ft from the exterior wall of the DOME Temperature Reactors may have operating temperatures up to 800C The RSS shall be capable of removing up to 100 KW of heat during normal operating conditions the concrete structure has an upper temperature limit of 100°C

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system: Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system

Converting abundant lignocellulosic biomass to sugars as fungible precursors to fuels and chemicals has the potential to diversify the supply chain for those products, but further process improvements are needed to achieve economic viability. In the current work, process intensification of the key enzymatic hydrolysis unit operation is demonstrated by means of a membrane reactor system that was operated continuously. Lignocellulosic biomass (pretreated corn stover) and buffered enzyme solution were fed to a continuously stirred-tank reactor, and clarified sugar solution was withdrawn via a commercial tubular ultrafiltration membrane. The membrane permeance decline and membrane cleaning efficacy were studied and did not vary significantly when increasing fraction insoluble solids (FIS) from 2.5% to 5%. Continuous enzymatic hydrolysis was successfully operated for more than 80 h. A model for the reactor system was able to predict dynamic behavior that was in reasonable agreement with experimental results. The modeled technical performance of anticipated commercial batch and continuous enzymatic hydrolysis processes were compared and showed that continuous operation would provide at least twice the volumetric productivity for the conditions studied. Further improvements are anticipated by better membrane selection and by increasing FIS.

09 BIOMASS FUELS↗
DOE OSTI2017

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Advanced Reactor Technology Program Molten Salt Reactor Campaign FY 2018 Summary

In FY 2018, the inaugural Molten Salt Reactor (MSR) Campaign was begun with the US Department of Energy (DOE) Advanced Reactor Technology (ART) program. The campaign is led out of Oak Ridge National Laboratory (ORNL) under the direction of Lou Qualls, who serves as the national technical director for the program. The campaign is a revival and refocusing of a significant DOE research program from the 1950’s, 1960’s and 1970’s, which culminated in the operation of the Molten Salt Reactor Experiment. The principal objectives of the campaign’s first year were to study and define the current condition of the technology, identify and prioritize technology gaps, and establish a national team of experts to collaboratively address the technical needs. The highest-level objective of the campaign is to facilitate the development and near-term market deployment of MSRs in conjunction with industry partners.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

First Annual Report on Pulsed Thermal Tomography Nondestructive Evaluation of Additively Manufactured Reactor Materials and Components: Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components

Additive manufacturing (AM, or 3D printing) for commercial nuclear energy applications is an emerging method for cost-efficient manufacturing aimed at replacing aging nuclear reactor parts and reducing costs for new construction. Because of the geometry of metallic structures of interest for nuclear applications, which consist of planar primitives with no symmetry of revolution, limited options are available for nondestructive evaluation (NDE), either during or post manufacturing. Known material flaws in AM include low-density regions consisting of non-sintered powder, which have to be detected to ensure the safety of long-term performance nuclear reactor components. As a solution to NDE of AM, we are developing pulsed thermal tomography (PTT) models and depth inversion algorithms for 3D imaging. PTT has many advantages because the method is non-contact and allows for in-service, NDE of AM nuclear reactor parts. By analyzing transients of surface temperature response due to internal thermal resistances, one can obtain 3D reconstructions of material effusivity using a unique inversion algorithm developed at Argonne. This report provides results of preliminary performance evaluation of PTT capability in detection of flaws in metallic structures, as well imaging complex geometry AM structures. Performance of PTT was investigated through modeling and experiments, Computer models of PTT were developed using COMSOL Heat Transfer module. Defects were modeled as cylindrical flat bottom holes (FBH), which is a common model of calibrated material flaws in thermal tomography experiments. Materials considered in this study include stainless steel 316 (SS316), stainless steel 304 (SS304), and Inconel 718. Theoretical analyses were conducted to validate inversion of simulated PTT data with COMSOL for a plate. Subsequently, 3D reconstructions were performed on COMSOL simulations for FBH, revealing a decrease in spatial resolution over depth due to thermal diffusion. The results of this study show that the performance of the inversion algorithm for detecting smaller defects depends strongly on the depth of the defect as well as the incident heat flux. The size of detectable defect was estimated by fitting a Gaussian function to surface temperature profile. The criteria for detectability was taken as 20mK noise equivalent temperature difference (NETD), which is currently the sensitivity limit of high-performance infrared cameras. It was determined through computer simulations that the smallest detectable FBH in SS316 has a 50µm diameter and is located 0.5mm below the plate surface. Preliminary experimental studies were performed to evaluate PTT performance in detecting calibrated flaws using metallic specimens with AM defects simulated as drilled-in FBH’s. High strength Stainless Steel 316 and Inconel 718 alloys were considered, as well as lower grade Stainless Steel 304, Nickel 200, and Hastelloy C276. Specimens investigated in this report consisted of approximately 1/4in-thick plates made out of these alloys using conventional manufacturing methods. The diameters of FBH’s varied from 1mm to 8mm, and their depths below the plate flat surface varied between 1mm and 6mm. The size of the smallest FBH was limited to 1mm because conventional mechanical drills were used for creating the holes. PTT imaging results have shown that 1mm-diameter FBH located 1mm and 2mm below the surface were detectable. Larger size FBH were detectable at greater depth. For example, 6mm-diameter FBH could be detected at 8mm depth. Image contrast varied slightly between the specimens, with the best reconstructions obtained in SS316 and C276 plates. Finally, PTT capability in imaging several AM structures fabricated from Inconel 718 (IN718) powder feedstock was investigated. The AM structures have complex geometry, but do not have calibrated internal defects. The objective is to determine PTT settings, such as the total integration time, for imaging of representative AM structures. A new FLIR X8500sc IR camera with higher spatial resolution (1280x1024 at 181Hz frame rate) was integrated into the PTT laboratory setup. Software tools were developed to ensure compatibility of new camera files with legacy effusivity reconstruction codes. The new camera was used to obtain 3-D reconstructions of AM IN718 nozzle plate and a particle filter plate. It was shown that PTT can scan through 2/3in-thick Inconel 718 plates in approximately 15s. Several modes of 3-D data visualization were explored, including using ImageJ and MATLAB software packages. The results demonstrate the capability of PTT in 3-D imaging of complex geometry AM structures. Intermediate results demonstrated in this report will be developed further to investigate detection of flaws in AM specimens.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Grizzly Development for Light Water Reactor and Advanced Reactor Applications in Fiscal Year 2021

This report summarizes work performed during Fiscal Year (FY) 2021 at Idaho National Laboratory (INL) for the U.S. Department of Energy’s Nuclear Engineering Advanced Modeling and Simulation (NEAMS) program for the Structural Materials and Chemistry Technical Area in the work package entitled “MS-21IN050106 - Structural Materials - INL.” This effort mainly focused on development and application of capabilities for engineering-scale analysis of nuclear reactor structural components in the Grizzly and BlackBear codes. These efforts include performance improvements and application of models for high-temperature component response, cluster dynamics modeling of precipitation in light water reactor pressure vessel steel, engineering-scale capability development in the areas of concrete simulation and reactor pressure vessel analysis, and preparation for and conducting an independent assessment of the adherence of these codes to software quality standards.

36 MATERIALS SCIENCE↗