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At least 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Development and Rigorous Validation of Modules for Predicting Organic Aerosol CCN Activity from Molecular Properties

The indirect interaction of anthropogenic aerosols with clouds (aerosol indirect forcing) represents one of the main uncertainties in climate projections, and reducing this uncertainty requires improved understanding of the aerosol life cycle, including atmospheric aging and the ability to nucleate cloud droplets or crystals. A key parameter that links the chemical properties of aerosol to its nucleating properties, and ultimately to aerosol indirect forcing, is its hygroscopicity (ability to uptake water). One way to compare and systematize the ability of various chemical compounds or particle types to nucleate cloud drops is via the hygroscopicity parameter κ, which can be used directly in global models to simulate aerosol indirect forcing. In this project we further investigated the relationships between the chemical composition of organic aerosol particles, which are known to be important components of atmospheric particulate matter, and their ability to serve as cloud condensation nuclei (CCN). In particular, we isolated the role of viscosity on κ by applying methods similar to those we used previously in our DOE/ASR research. Our motivation for the focus on viscosity was through its link to particle phase state. Particle phase affects water uptake on timescales relevant to the laboratory and also likely relevant to atmospheric processes, with reduced apparent hygroscopicities for semisolid and solid particles, compared with liquid particles. Water uptake itself affects viscosity, an important feedback that we explored in this project. The objectives of the project were as follows: (1) measure the effects of molecular structure, temperature, and relative humidity on the viscosity of organic aerosol formed via controlled laboratory experiments, (2) measure the effects of organic aerosol viscosity on water uptake and CCN activity above and below water saturation, (3) further develop the κ parameterization of organic aerosol hygroscopicity to include viscosity/phase, and (4) experimentally validate the κ parameterization for organic aerosol with chemical composition ranging from relatively simple to complex. It was proposed that particle viscosity exerts a kinetic control on equilibrium water uptake. High viscosity was expected to delay water uptake based on diffusional limitations. This would manifest itself in reduced observed CCN activity or κ. Understanding the interplay between water content and viscosity (composition control on particle viscosity) and between viscosity and equilibrium water content (diffusional control on water uptake) were the overall science objectives of this grant. Targeted experiments were carried out to test the kinetic control hypothesis. These experiments included studies that explicitly measured the equilibration time scale of water uptake, and experiments that tested the role of viscous particle coatings on the observed CCN activity. From the combined work, it is concluded that mass transfer limitations by glassy organic particles are unlikely to affect cloud droplet activation near laboratory temperatures. This implies that viscosity does not need to be included in parameterizations that link particle chemical composition and κ. A parameterization to predict κ through the UNIFAC group contribution method was developed with previous ASR funding. As planned, this computational module was further developed and validated here through targeted laboratory studies. The computational module has been made available as free software under the GNU public license. A database of κ-values to aid model development has been compiled and is available online via an open access model. In conducting this work, significant effort was made to understand how aerosol composition, including the associated aerosol water content, affects viscosity. This included a study detailing the influence of functional group composition on organic aerosol viscosity. We identified a strong link between volatility and viscosity. Functional groups responsible for low vapor pressure also promote high viscosity. Specifically, carboxylic acid and hydroxyl groups were identified as major contributors to high viscosity. Several studies were carried out to characterize the link between composition and the glass-to-liquid phase transition. Compositions tested include models of organic/inorganic particles and models of complex organic aerosol formed through secondary processes. A simple computational model to predict the amorphous phase diagram was developed and tested. Excellent agreement between model and measurements was obtained for organic/inorganic mixtures. The computational model is simple enough to be included in global model simulations. However, constraints of input parameters such as the Gordon-Taylor constant for binary mixtures for a wide range of complex mixtures have not yet been fully evaluated. Combined, the performed work provides new constraints on the amorphous phase state of organic aerosols.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 6: Surrogate Sample Preparation and Validation Experiment Test Plan

This document presents the experimental test plan for Task 6: Surrogate Preparation and Validation. Task 6 will produce and validate surrogate oxyhydroxide materials used in other experimental tasks within the larger Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Project. Task 6 includes both the preparation and characterization of surrogate oxyhydroxide materials as well as the analysis of oxide/oxyhydroxide layers on Savannah River Site Spent Nuclear Fuel Materials. This test plan provides experimental design details and the task’s overall quality assurance approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Optimization of Advanced Steels for Cyclic Operation through an Integration of Material Testing, Modeling and Novel Component Test Validation

The efforts in this project were focused in three primary objectives, provided below: • To develop the needed microstructural processing and performance relationships and associated material models for specific constituents in fabricated weldments (such as the parent material, heat affected zone regions and weld metal), • Apply these metallurgical relationships through modeling of a welded pressure bearing power plant component subjected to cyclic operational conditions under both mechanical and thermal loading, and • Validate the model through novel structural feature and component tests. An attempt was made to optimize the heat treatment schedule to reduce or eliminate the presence of boron nitride (BN) in the matrix of the Grade 92 material. BN has been shown to be a primary nucleation source for creep damage. Proper evaluation of material behavior, and as assessed in this research, included the influence of geometric notches (such as those from poor fabrication or design) and metallurgical notches (such as weldments or poorly processed parent material). Better practice methodologies were implemented in the theoretical design of a high-temperature, critical component for cyclic operation. It was demonstrated that vessel testing can be reliably performed and provides a critical link between standard testing, structural assessment and validation. While the goal of an improved Grade 92 steel was not achieved in this project, there is considerable value in applying the documented testing and assessment methodologies to assess material and/or component behavior. This is especially true for materials and/or designs which are expected to be subjected to flexible operating modes. Design of a cycling state of the art thick wall steam header for long term operation with a state of the art ferritic steel was a task in this project. Implementation of that task showed how many factors including plant operating mode, creep strength, creep ductility, fatigue strength, and resistance to crack growth should be considered and evaluated to properly design high temperature components for long term service (200,000 operating hours or more) in a modern Fossil energy power plant. The current ASME non-nuclear boiler design code does not mandate this level of complex analysis nor does it provide all the metallurgical and mechanical properties data needed to perform this kind of design analysis.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2018

Quick Urban and Industrial Complex (QUIC) CBR Plume Modeling System: Validation-Study Document

This document contains information on QUIC validation studies that have been performed to evaluate the wind, pressure, and concentration solutions produced by QUIC. Tests of QUIC's source term models and specialized transport mechanisms (e.g., buoyancy algorithms, particle settling and deposition, liquid pool growth, droplet and pool evaporation) are also provided. A brief description of the experiment, a few example validation results, and references to journal articles, reports, and/or presentations are given. Both LANL and non-LANL sources are provided.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Energy Storage System Safety Documenting and Validating Compliance with Codes and Standards

Energy, environmental, and economic challenges are spurring more widespread consideration and use of energy storage systems (ESSs), which in turn are driving increased development of new ways to store energy electrochemically, mechanically, and thermally. These new methods necessitate an increased focus on ensuring that public health, safety, and welfare are not adversely affected—something that has been addressed for many years through codes, standards, and regulations (CSR5)1. CSRs provide requirements that establish a basis for determining if an ESS is safe, whether electrochemical, mechanical, or thermal and regardless of the range of ESS applications, energy capacities, physical sizes, location, or number installed at any given site. The key to achieving desired safety goals, as memorialized through CSRs, is through documenting and validating compliance with applicable CSRs. The process of documenting and validating compliance, which is a key component to the initial approval as well as continuing acceptance of an ESS installation, is generally called conformity assessment.

25 ENERGY STORAGE↗
DOE OSTI2019

Modeling and Validation of Irradiation Damage in Ni-based Alloys for Long-Term LWR Applications (Final Report)

As light water reactor (LWR) plant lives are extended, the need for predictive modeling tools for materials degradation increase in order to ensure safe operation and plan for component replacements. The Grizzly code, which is built on the MOOSE multi-physics simulation environment, is being developed for exactly this purpose. To extend Grizzly to include capabilities for modeling Ni-based alloys, we propose a US/UK integrated program to address thermal and irradiation-induced transformations mechanisms of Alloys 690 and 625. Alloy 690 is widely used in existing LWR plants due to its superior stress corrosion cracking (SCC) resistance compared to Alloy 600. Alloy 625 is used in more limited applications but offers the benefits of both high strength (in the aged condition) and corrosion/SCC resistance. Research has shown that both alloys can undergo phase changes due to thermal or irradiation exposure. In the precipitation-hardened condition, Alloy 625 “softens” during neutron irradiation as the strengthening precipitates decompose and metastable precipitates form. However, the nature and rates of these transformations as a function of exposure conditions are not well understood. Similarly, the effects of these thermal and irradiation-induced microstructural changes on mechanical properties require evaluation. The proposed program combines thermal and irradiation experiments, mechanical testing, microstructural characterization using state-of-the art analytical techniques, atomistic modeling, micro-and macro-scale modeling via Grizzly. This multi-pronged approach yields several benefits: 1) atomistic scale models capture the physics of the mechanisms of phase transformations, which can be generalized to other systems, 2) the modeling results (from atomistic to continuum) will be validated experimentally to ensure the predictive aspect, 3) the microstructure/property relationships captured with mechanical testing provides information at the engineering scale that is needed to validate the integration of the proposed new capability into Grizzly

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Field Validation of High-R Windows: Market Assessment and Program Plan

This report provides a market assessment for high-performance residential windows based on information gathered from key researchers and industry stakeholders including a series of interviews conducted with high-performance builders in 2019. The report also outlines the U.S. Department of Energy’s planned field validation activities in support of high-R windows and provides the basis for planned field validation approaches. Although the report may be of general interest to those who work in fields pertaining to building energy efficiency and/or market adoption of energy-efficiency measures, the primary intent of this report is to document market characteristics, barriers, and activities related to high-performance residential windows in order to assist with internal program planning.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Validation of Sodium Chemistry (NAC) Package - FY18 Progress.

This report describes the progress on the validation of the development of MELCOR Sodium Chemistry (NAC) package. The primary focus for this report is to ensure that the implementation of the CONTAIN-LMR sodium models into MELCOR is correctly done. Thus, the verification test is to conduct the code-to-code comparison with MELCOR and CONTAIN-LMR. Last year we had reported the development of NAC package which included three sodium models: spray fire, pool fire and atmospheric chemistry. The first 2 models were completed and additional improvement for these two models were done this year to allow upward spray capability and various functional capability for modeling the pool fire experiment better, respectively. This year, the atmospheric chemistry implementation has been progressed to a point for testing in the presence of water vapor (modeled as ideal gas) as a part of the two-condensable option model in the CONTAIN- LMR. The user's guide and reference manual for the NAC package including these improvements are described in a separate document being published as a part of the MELCOR 2.2 release. For this report, we would discuss the experimental validation using the implemented spray fire and pool fire models. A code-to-code comparison with CONTAIN-LMR is described for a spray fire experiment. Note that the atmospheric chemistry model has not fully implemented due to the absence of the two condensable option. Only the chemical reactions between the sodium aerosol and water vapor can be modeled. ACKNOWLEDGEMENTS This work was overseen and managed by Matthew R. Denman (Sandia National Laboratories). In addition, we appreciate that Chris Faucett for developing experimental data and provided the initial input decks as a part of the MELCOR assessment report development for U.S. Nuclear Regulatory Commission's project. This work is supported by the Office of Nuclear Energy of the U.S. Department of Energy work package number AT-17SN170204 and NT-185N05030102.

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

OSIRIS Validation and Verification Report

The On-Site Inspection RadioIsotopic Spectroscopy (OSIRIS) system is a gamma-ray spectroscopy instrument for on-site inspections conducted under the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The OSIRIS software filters spectral analysis data, revealing information to users on only 17 CTBT-relevant fission-product isotopes. In this report, we review the CTBT requirements for on-site inspection radionuclide identifiers, plus additional requirements imposed by the design team. Then a series of validation tests conducted during OSIRIS development are described. The validation tests were designed to gauge if the nascent OSIRIS system met those requirements. We conclude with a set of recommended verification tests to ensure that newly-manufactured OSIRIS systems also will meet the requirements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Separate Effects Validation for Subcooled Boiling in CTF

In 2010, the U.S. Department of Energy created its first Energy Innovation Hub, which is focused on developing high-fidelity and high-resolution modeling and simulation (M&S) tools for modeling of light water reactors (LWRs). This hub, the Consortium for Advanced Simulation of LWRs (CASL), has developed an LWR simulation tool called the Virtual Environment for Reactor Applications (VERA). The multi-physics capability of VERA is achieved through the coupling of single-physics codes, including CTF (the CASL version of Coolant Boiling in Rod Arrays— Three Field (COBRA-TF)), Michigan Parallel Characteristics Transport (MPACT), BISON, and Materials Performance and Optimization (MPO) Advanced Model for Boron Analysis (MAMBA). As part of its M&S efforts, CASL has identified various challenge problems, including Crud Induced Power Shift (CIPS), Crud-Induced Localized Corrosion (CILC), Pellet-Cladding Interaction (PCI), and Departure from Nucleate Boiling (DNB). This work addresses CASL milestone L2:VVI.P19.03, which focuses on uncertainty quantification of crud, which is relevant to both CIPS and CILC. This is achieved through an analysis and separate effects validation of the thermal hydraulic phenomenon known as subcooled boiling. As part of this work, various sources of experimental data are examined and compared to different options for empirical modeling of subcooled boiling. Through this analysis, a complete understanding of the underlying models and their implementation details are understood. A subset of these data are incorporated into a separate effects validation study of CTF. The Westinghouse Advanced Loop Tester (WALT) and Rohsenow experiments are modeled, and it is shown that the newly-implemented Gorenflo correlation is more accurate than the existing Chen and Thom correlations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Verification and Validation of the STAT7 Code

This document presents the verification and validation of the STAT7 Version 1.0 code which has been developed to perform the steady-state thermal hydraulic safety calculation of the MIT Research Reactor required upon conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel. The key capabilites of the code including the statistical processing, water property generation, and thermal hydraulic solution are verified independently. While the first two capabilites are checked using statistical tests and NIST data tables, each component of the thermal hydraulic solution is verified using simple geometry test problems. The comparison between results obtained by the code and hand calculations on these problems shows that each of these components work correctly. Additionally, a comprehensive comparison between STAT7 Version 1.0 and PLTEMP/ANL Version 4.2 calculations for a five-channel test problem is performed for the validation. The good agreement between the obtained results for all thermal hydraulic parameters belonging to the fuel core region is achieved. An issue in bypass flowrate calculation was detected when both fin-friction and bypass flow claculations are involved. This problem can be corrected in a later version of the code, however, does not affect the application of STAT7 code as it has been used to date where the flow distribution in the core region and bypass channel is specified as an input parameter.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Involute Working Group – Validation of CFD Turbulence Models for Steady-State Safety Analysis

There are three research reactors in the world having fuel plates curved as circle-involute (a spiral generated around a circle): The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) located in Tennessee, U.S.A. [ORNL, 2018]; The Laue-Langevin Institute (ILL) High Flux Reactor (RHF) located in Grenoble, France [ILL, 2018]; and, The Technical University of Munich (TUM) Research Neutron Source Heinz Maier-Leibnitz (FRM II) located in Garching, Germany [TUM, 2018]. All three reactors are currently using Highly Enriched Uranium (HEU, 235U/U ≥ 20 wt. %) as fuel and all three are actively engaged in activities to convert to Low Enriched Uranium (LEU, 235U/U < 20 wt. %) fuel. More information on these reactors can be found in Section two. For various reasons, these reactors have expressed interest in using Computational Fluid Dynamics (CFD) tools to perform their Steady-State Thermal-Hydraulic (SSTH) safety calculations. Using CFD tools for SSTH analysis represents generally a significant departure from traditional methods and requires thorough verification and validation (V&V) to be accepted by regulators. This is why Argonne National Laboratory (ANL) and the involute reactors formed an informal group to help each other in this endeavor. This so-called Involute Working Group (IWG) aims at supporting the qualification of CFD tools for SSTH safety calculations. Activities include benchmarking, code-to-code comparison, Verification and Validation (V&V) as well as technical support. More information on the IWG and the reasons why involute-plate reactors are interested in CFD can be found in Section three. The present study aims at testing three turbulence models commonly implemented in commercial software (namely K-Epsilon, K-Omega and Reynolds-Stress) and assessing their relevance for involute-plate reactor application. An experiment performed by Gambill and Bundy in 1961 was performed to measure heat transfer coefficients in conditions prototypic of the HFIR reactor [Gambill, 1961]. More details on this experiment are provided in Section four. Based on the information available, CFD models of this experiment have been created and the experimental results have been compared to the ones obtained with these CFD models. CFD models and the results of this comparison are provided in Section five.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

ATF material model development and validation for priority cladding concepts

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the Bison fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on the focus was on U 3 Si 2 fuel and FeCrAl cladding as part of a high impact problem through the Nuclear Energy and Advanced Modeling Simulation (NEAMS) program. Then, developments for Cr 2 O 3 -doped UO 2 fuel, and SiC-SiC and Cr-coated zirconium-based claddings began based upon industry interests. In late fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors (CASL) took over further ATF work in Bison in support of the Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC identified their list of priority fuel and cladding concepts, which included Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and Cr-coated zirconium-based and FeCrAl claddings. In particular, the NRC suggested that reports similar in form to NUREG/CR-7024 that was developed for traditional LWR materials UO 2 and zirconium-based claddings (i.e., Zircaloy-4, M5 ® , ZIRLO TM ) be created for the priority ATF concepts. The approach to ATF capability development in Bison since the beginning has been two-fold: (1) empirical correlations and (2) multiscale model development. Both approaches have uncertainty inherent to them. Uncertainty in empirical correlations is bounded by the experimental data upon which with the correlation was developed. Models developed through a multiscale approach have uncertainty associated with the lower length scale calculations and input parameters that must be propagated to the engineering scale model in Bison. In this report, the recommended models, their range of applicability (e.g., temperature, fluence), and associated uncertainty for the NRC priority clad concepts Cr-coated zirconium based clad and FeCrAl are presented in a manner similar to the aforementioned NUREG. In addition, in the absence of experimental data for validation, demonstration cases are presented that investigate the ballooning and burst behavior of Cr-coated and FeCrAl clads as well as tritium permeation in FeCrAl. A portion of the study by Gamble et al. on FeCrAl failure is also revisited with the latest updated models. First, the nominal results are shown using the as implemented models. Then a sensitivity analysis and uncertainty quantification is performed to assess the impact of the uncertainties in the models on the parameters of interest. For the Cr-clad ballooning case the parameter of interest is the time to burst. For the revisited FeCrAl study, the parameter of interest is the pressure at burst.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Verification and Validation of the STAT7 Code

This document presents the verification and validation of the STAT7 Version 1.0 code which has been developed to perform the steady-state thermal hydraulic safety calculation of the MIT Research Reactor required upon conversion of the reactor from highly enriched uranium fuel to low- enriched uranium fuel. The key capabilities of the code including the statistical processing, water property generation, and thermal hydraulic solution are verified independently. While the first two capabilities are checked using statistical tests and NIST data tables, each component of the thermal hydraulic solution is verified using simple geometry test problems. The comparison between results obtained by the code and hand calculations on these problems shows that each of these components work correctly. Additionally, a comprehensive comparison between STAT7 Version 1.0 and PLTEMP/ANL Version 4.2 calculations for a five-channel test problem is performed for the validation. The good agreement between the obtained results for all thermal hydraulic parameters belonging to the fuel core region is achieved. An issue in bypass flowrate calculation was detected when both fin-friction and bypass flow calculations are involved. This problem can be corrected in a later version of the code. However, it does not affect the application of STAT7 as it has been used to date where the flow distribution in the core region and bypass channel is specified as an input parameter.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Experimentally Validated Computational Modeling of Creep and Creep-Cracking for Nuclear Concrete Structures

In a Nuclear Power Plant, one of the most important components is the concrete nuclear reactor cavity, which serves both a structural and protective function as the biological radiation shield. Given that creep has been identified as a major knowledge gap in the assessment of nuclear structures (NUREG/CR-7153), this work helps to further the understanding of creep behavior of massive concrete containment structures for decades to enable safe and long-term operation of these facilities. This project has developed a robust, experimentally validated model to predict creep in nuclear concrete structures for up 60 years using short-term creep data thereby enabling a longer service life of critical facilities and early detection of structural failure. The work presented in this report is a pairing of computational and experimental methods. For the first time, the time temperature superposition (TTS) principle was successfully used to generate a uniaxial creep compliance master curve to predict mortar creep response for up to 22,500 days (nearly 60 years) at a reference temperature of 20°C. These data were used as input into finite element analysis (FEA) codes that use highly realistic random, 3D concrete microstructures from reconstructed coarse limestone aggregates. Finite element analysis performed provides the ability to quickly upscale mortar viscoelastic behavior to long-term concrete creep/relaxation data. A master creep compliance curve, constructed from the TTS principle, spanning 27 years, was used to validate two and a half decades of simulated concrete creep. Concurrently, three different simulated wall specimens were designed to mimic the behavior of post-tensioned concrete nuclear containment facility vessel walls over time as a result of concrete creep. The specimens were designed with different thicknesses, transverse and longitudinal reinforcement ratios, and level of post-tensioning stress. Each specimen contained various instrumentation to measure internal concrete temperature, concrete strain, and post-tensioning strain hourly for over 3 years. The concrete creep model developed in this project, based on the FEA concrete simulations, was applied to simulate the structural-scale experiments of prestressed concrete walls conducted in this project using the Grizzly code. These models can represent the effects of reinforcing and prestressing. Although there are some discrepancies with the experimental data, the model can predict the overall trends of the creep response in these experiments. One of these experimental models was also applied to an extended time to demonstrate how the findings from this study can be used to predict the behavior of actual structures of interest that have been in service for extended periods of time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

FY20 SAM Code Developments and Validations for Transient Safety Analysis of Advanced non-LWRs

The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs. Rapid developments continued in fiscal year 2020 (FY20) to support various needs of the advanced reactor community, especially the NRC and industry on the licensing safety analysis of advanced reactor designs. Significant code changes were made to provide various capability enhancements, bug fixes, and user friendliness improvements. Major code updates are summarized in Section 1, while four important enhancements are detailed in Sections 2-5, including a multi-dimension flow model; reactivity feedback and decay heat models; control and trip system modeling, and additional fluid and solid thermophysical property models. Code validation activities in FY20 include using test data from the Fast Flux Test Facility (FFTF), the High Temperature Test Facility (HTTF), and several separate effects test facilities for pebble-bed modeling.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

HPAC Validation against NRC Documentation

This report describes validation of the radiological source terms in the Nuclear Facility Accident (NFAC) module of the Defense Threat Reduction Agency (DTRA)’s Hazard Prediction and Assessment Capability (HPAC) software. The validation will be conducted by comparing the source terms defined in HPAC to published data from references approved by the U. S. Nuclear Regulatory Commission (NRC). Comparisons will be done for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). This comparison includes the magnitude, isotopic composition, and timing of the release fractions over the course of a severe or moderate accident. Calculations for both facility-specific nuclear power plants and for more generic reactor categories are involved. It is shown that the NFAC radiological source terms in HPAC agree well with NRC standards for moderate and severe accidents at PWRs and BWRs.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗