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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

VERA-CS Verification & Validation Plan

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISON / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

File-Based One-Way BISON Coupling Through VERA: User's Manual

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention [1–6]. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code [8,9] being developed by the Idaho National Laboratory (INL) is the focus of ongoing work in the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework [10], BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, this work uses a 2D azimuthally symmetric (R-Z) smeared-pellet model. This manual is intended to cover (1) the procedure pertaining to the standalone BISON one-way coupling from VERA and (2) the procedure to generate BISON fuel temperature tables that VERA can use.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Final Technical Report: Imaging a Dry Storage Cask with Cosmic Ray Muons

The goal of this project is to build a scaled prototype system for monitoring used nuclear fuel (UNF) dry storage casks (DSCs) through cosmic ray muon imaging. Such a system will have the capability of verifying the content inside a DSC without opening it. Because of the growth of the nuclear power industry in the U.S. and the policy decision to ban reprocessing of commercial UNF, the used fuel inventory at commercial reactor sites has been increasing. Currently, UNF needs to be moved to independent spent fuel storage installations (ISFSIs), as its inventory approaches the limit on capacity of on-site wet storage. Thereafter, the fuel will be placed in shipping containers to be transferred to a final disposal site. The ISFSIs were initially licensed as temporary facilities for ~20-yr periods. Given the cancellation of the Yucca mountain project and no clear path forward, extended dry-cask storage (~100 yr.) at ISFSIs is very likely. From the point of view of nuclear material protection, accountability and control technologies (MPACT) campaign, it is important to ensure that special nuclear material (SNM) in UNF is not stolen or diverted from civilian facilities for other use during the extended storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

CASL Verification and Validation Plan

This report documents the Consortium for Advanced Simulation of LWRs (CASL) verification and validation plan. The document builds upon input from CASL subject matter experts, most notably the CASL Challenge Problem Product Integrators, CASL Focus Area leaders, and CASL code development and assessment teams. This document will be a living document that will track progress on CASL to do verification and validation for both the CASL codes (including MPACT, CTF, BISON, MAMBA) and for the CASL challenge problems (CIPS, PCI, DNB). The CASL codes and the CASL challenge problems are at differing levels of maturity with respect to validation and verification. The gap analysis will summarize additional work that needs to be done. Additional VVUQ work will be done as resources permit. This report is prepared for the Department of Energy’s (DOE’s) CASL program in support of milestone CASL.P13.02.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

2017 NE-4 Nuclear Technologies Research and Development (NTRD) Achievements Report

Fiscal year (FY) 2017 initiated new and reestablished presidential and departmental nuclear energy priorities for the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE) that will: • Support the Administration’s priorities to achieve U.S. energy dominance and economic prosperity through U.S. technology leadership • Support DOE-NE mission priorities and goals for the existing nuclear fleet, advanced reactor pipeline and national fuel cycle infrastructure • Focus on early stage research, and effective cooperation with industry, academia, and other government agencies • Propose investments to maintain and improve capabilities and infrastructure to conduct leading-edge science, research, and technology development. DOE-NE’s mission priorities will focus on the existing nuclear fleet, advanced reactor pipeline, and fuel cycle infrastructure. Therefore, the DOE Nuclear Technology Research and Development (NE-4) mission aligned to (1) support the research and development (R&D) of innovative reactor technologies that may offer improved safety, functionality, and affordability and (2) conduct research to reduce long-term technical barriers for advanced nuclear energy systems, including space fission systems. Key activities to execute this mission include: • To achieve the NE-4 mission priorities, two major programs with eight R&D campaigns are engaged in impactful research, development, demonstration, and deployment: Fuel Cycle Technologies (FCT) Program • Fuel Cycle Options Campaign evaluates complex fuel cycle options and provides performance metrics on various systems and scenarios to improve the knowledge of interdependencies between technologies and different systems. • Advanced Fuels Campaign performs science-based R&D on accident tolerant fuels in addition to advanced light-water reactor (LWR) fuel and transmutation fuel. • Joint Fuel Cycle Studies (JFCS) is a collaboration with the Republic of Korea to assess the feasibility and nonproliferation of electrochemical recycling to manage used fuel. • Material Recovery and Waste Form Development (MRWFD) Campaign researches advanced fuel cycle material recovery and waste management capabilities to improve fuel cycle performance with less processing and waste generation. • Material Protection, Control, and Accountability Technologies (MPACT) Program develops tools and capabilities to secure the next generation nuclear materials management and safeguards for nuclear fuel cycles. Advanced Reactor Technologies (ART) Program • Fast Reactors Campaign targets benefits in either resource utilization or waste management and possesses favorable features for small reactor applications and plutonium management. • Gas-Cooled Reactors Campaign seeks to raise the technological readiness of high temperature gas-cooled reactors (HTGRs) and other reactor concepts utilizing coated particle fuel, graphite, or helium coolant by performing high risk, enabling research and development that commercial developers are unable or unwilling to undertake in the current energy market. • ART Energy Conversion (EC) Project provides solutions to convert heat from an advanced reactor to useful products that support commercial application of the reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

NTRD Technical Monthly October FY18

Topics covered in this Nuclear Technology and Research Development (NTRD) monthly update include: Advanced Fuels Campaign, Material Recovery and Waste Forms Development, MPACT Campaign, Fuel Cycle Options Campaign, Joint Fuel Cycle Study Activities, and AFCI-HQ Program Support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Evaluation of Solver and Time Stepping Options and RIA Demonstration

This report summarizes Bison development and application in three areas: • Development of the layered 1D approach to permit its use with the reference residual convergence criterion that is widely used for 2D and 3D Bison simulations. For higher-dimensional simulations, this approach has resulted in significant improvements in robustness and solution efficiency, and it is expected to result in similar benefits for 1D simulations of light water reactor (LWR) rods. • A study on the effects of the chosen convergence tolerances on solution difficulty and accuracy was performed for a prototypical layered 1D simulation of an LWR fuel rod. The effect of convergence tolerance on iteration counts, run times, and several quantities of interest in the solution are compared for the various convergence tolerance. This study will be helpful for guiding the choice of convergence tolerances. • A demonstration of a reactivity insertion accident (RIA) simulation with Bison within the coupled VERA code was performed. The MPACT neutron transport code and the CTF thermal-hydraulic code provided inputs to Bison for a postulated control rod ejection scenario. This demonstrates the ability of the VERA environment to consider coupling effects not only for normal operation, but for accident conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Nuclear Technology Research and Development (NTRD) Technical Monthly June FY18

The Nuclear Technology Research and Development (NTRD) Technical Monthly Report for June FY18 covers the following areas: Advanced Fuels Campaign, Material Recovery and Waste Forms Development, MPACT Campaign, Fuel Cycle Options Campaign, Joint Fuel Cycle Study Activities, and, AFCI-HQ Program Support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Review of V&V Documentation

Consortium for Advance Simulation of Light Water Reactors (CASL) is a Department of Energy Innovation Hub whose mission is the following, "CASL is a collaboration of the nation's leading scientists, institutions, and supercomputers, with an aggressive 10-year mission to confidently predict the performance of existing and next-generation commercial nuclear reactors through comprehensive, science-based modeling and simulation." The CASL program to date has focused on developing the necessary predictive capability. Rightly so, it is characterized by many as a research project. As a matter of intent, the first 6 years of CASL focused on developing and demonstrating the prediction capability of a suite of independent physics codes: MPACT (neutronics), CTH (thermal hydraulics in the core), BISON (fuel performance), and MAMBA (CRUD and boron uptake on fuel rod surfaces) The last 4 years focused on initial attempts to couple the codes and to demonstrate those capabilities through a series of challenge problems aligned to 3 key issues of interest to the nuclear power industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of a Pyroprocessing Flowsheet Model for Safeguards Analysis

Pyroprocessing flowsheets have been designed that have efficient actinide recovery, meet nonproliferation objectives, minimize waste, and are commercially viable. The analysis of safeguards of pyrochemical recycle systems for the treatment of used fast reactor metal fuel can be greatly enhanced with the employment of robust dynamic models. Electrochemical process modeling also contributes to broader objectives of the MPACT program by supporting MC&A, options screening, and safeguards analysis and can inform development of sensor systems for a used fuel recycle facility. Modeling can also provide a method for defining metrics and requirements on advanced analysis techniques.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of a Liquid-Fueled Molten Salt Reactor Safeguards Model.

This work describes the ongoing work to develop a molten salt reactor (MSR) model and associated tools for safeguards analysis. A new flowsheet was developed in collaboration with Oak Ridge National Laboratory (ORNL) for the Molten Salt Demonstration Reactor (MSDR). This design was chosen by ORNL as a generic baseline design that could be used for safeguards research. The model has simple chemical processing that is less extensive than the two-fluid flowsheet developed in the last year. A detailed TRITON reactor physics model, provided by ORNL, was implemented into the process model. The process model now includes reactor parameters such as K-eff and decay heat, which could be used as part of an advanced safeguards approach. Finally, a set of generic safeguards tools based on current safeguards approaches were developed. These tools are flexible and can be used with most MSR flowsheets. ACKNOWLEDGEMENTS This work was funded by the Materials Protection Accounting and Control Technologies (MPACT) working group as part of the Fuel Cycle Technologies Program under the U.S. Department of Energy, Office of Nuclear Energy. The authors would also like to acknowledge Ben Betz ler for his work on the reactor physics models that were incorporated into the work and the continued collaboration with ORNL staff.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Modeling of Pyroprocessing Hot Cell for Process Monitoring Evaluation

In this report, we examine studies undertaken to determine the applicability of applying measurements with a neutron counter within a pyroprocessing hot cell. The studies focus on the modeling of the hot cell and neutron counter in Monte Carlo N- Particle Code (MCNP) to determine expected radiation effects and applicability of neutron counters. The neutron counter of focus was the High Dose Neutron Detector (HDND) developed at Los Alamos National Laboratory (LANL). We studied the HDND to determine what would be the detector response under normal background conditions with operating process equipment as well as the background’s potential effect on counting applications involving a U/TRU ingot produced by the pyroprocessing electrorefiner. To perform this task, we were required to develop models of the pyroprocessing process cell and unit operations and generate source terms for various different times in the process. The data for the pyroprocessing materials came from SNL’s Safeguards and Security Process Model (SSPM) as part of MPACT’s advanced integration mission. This report will discuss the development of the pyroprocessing process cell and unit operations models in MCNP as well as various simulations to determine the applicability of the HDND for process monitoring efforts in the hot cell.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly June FY-19

The monthly technical report for the Nuclear Fuel Cycle and Supply Chain (NFCSC) for June FY19 is provided. Major sections include: Advanced fuels campaign; Material recovery and waste forms development; MPACT campaign; Systems analysis and integration campaign; Joint fuel cycle study activities; and, AFCI-HQ program support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly October FY-20

The October FY20 Nuclear Fuel Cycle and Supply Chain (NFCSC) technical monthly report is provided. The usual topics are included: Advanced Fuels Campaign, Material Recovery and Waste Forms Development, MPACT Campaign, Systems Analysis and Integration Campaign, Joint Fuel Cycle Study Activities, and AFCI-HQ program Support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

FY20 Verification of BISON Using Analytic and Manufactured Solutions

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, Consortium for Advanced Simulation of LWRs (CASL), has developed an LWR simulation tool called Virtual Environment for Reactor Applications (VERA). The multi-physics capability of VERA is achieved through the coupling of single-physics codes, including BISON, CTF, MPACT, and MAMBA. BISON is a fuel performance code which models the thermo-mechanical behavior of nuclear fuel using high performance M&S. It is capable of modeling traditional LWR fuel rods, fuel plates, and TRi-structural ISOtropic (TRISO) fuel particles. It can employ three-dimensional Cartesian, two-dimensional axisymmetric cylindrical, or one-dimensional radial spherical geometry. It includes empirical models for a large variety of fuel physics: temperature- and burnup-dependent thermal properties, fuel swelling and densification, fission gas production, cladding creep, fracture, cladding plasticity, and gap/plenum models. This document details a series of code verification test problems that are used to test BISON. These problems add confidence that the BISON code is a faithful representation of its underlying mathematical model. The suite of verification tests are mapped to the underlying conservation equations solved by the code: heat conduction, mechanics, and species conservation. Twenty-two problems are added for the heat conduction solution, two for the mechanics solution, and none for species conservation. Method of Manufactured Solutions (MMS) capability is demonstrated with three problems, and temperature drops across the fuel gap are tested.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2019

Impact of Control Blade Insertion on the Deformation Behavior of SiC-SiC Channel Boxes in Boiling Water Reactors

This report describes the analysis of distortion of a Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composite channel box under in-reactor conditions of a boiling water reactor (BWR). The BWR core has significant gradients in the fast neutron flux across the channel box due to the presence of water rods within the fuel assemblies, and these gradients increase further with the insertion of control blades. As a result of temperature and neutron flux dependent irradiation induced swelling of SiC, the SiC-SiC composite channel box can undergo significant distortion. In this work, we evaluate the SiC-SiC channel box distortion for three different control blade positions. This analysis is based on the neutron flux and temperature distributions in the BWR core calculated using the neutronics code MPACT and thermalhydraulics code CTF. This calculation is coupled through temperature feedback. Subsequently, we have performed structural analysis based on the calculated neutron flux and temperature distributions, to determine the deformation and stress development in the channel box. The structural analysis was performed using the fuel performance modeling code BISON and the commercial finite element analysis software Abaqus. The results indicate that the channel box will undergo time dependent lateral bowing for all the control blade positions in the assembly. The bowing behavior is dominated by the swelling of SiC-SiC material under non-uniform neutron flux, and changes with variation in the control blade position. The lateral bending will cause a temporary interference between the channel box and the control blade. The developed stresses exceed the proportional limit stress of the material, which may cause matrix microcracking in the channel box. However, the stresses remain below the ultimate tensile strength of the material, and therefore, development of a full, through thickness crack in the channel box is not expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗