Assessment of SAMPSON Severe Accident Code Modeling against MAAP-MELCOR Crosswalk.
Abstract not provided.
Itoh, Ayumi↗
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OECD-NEA HTTF Benchmark Progress and Updates
General slides discussing High Temperature Test Facility benchmark progress, Prismatic HTGR deployment, modeling and simulation tools for validating systems, and verification and validation issues. Also discuss collaboration and conclusions of RELAP5-3D validation activities based on HTTF with trends in data and international impact to accelerate deployment of prismatic HTGR microreactors by providing an opportunity for designers to assess their codes against experimental data and solutions from other codes.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Verification and Validation Assessment for VERA
The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while capability refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Updated Verification and Validation Assessment for VERA
The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while credibility refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Updated Verification and Validation Assessment for VERA
The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while credibility refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
ASME Graphite code considerations
This summarizes the INL assessment of the ASME Code for graphite
36 MATERIALS SCIENCE↗
Impact of Corrections to the Spallings Volume Calculation on Waste Isolation Pilot Plant Performance Assessment
The numerical code DRSPALL (from direct release spallings) is written to calculate the volume of Waste Isolation Pilot Plant solid waste subject to material failure and transport to the surface (i.e., spallings) as a result of a hypothetical future inadvertent drilling intrusion into the repository. An error in the implementation of the DRSPALL finite difference equations was discovered and documented in a software problem report in accordance with the quality assurance procedure for software requirements. This paper describes the corrections to DRSPALL and documents the impact of the new spallings data from the modified DRSPALL on previous performance assessment calculations. Updated performance assessments result in more simulations with spallings, which generally translates to an increase in spallings releases to the accessible environment. Total normalized radionuclide releases using the modified DRSPALL data were determined by forming the summation of releases across each potential release pathway, namely borehole cuttings and cavings releases, spallings releases, direct brine releases, and transport releases. Because spallings releases are not a major contributor to the total releases, the updated performance assessment calculations of overall mean complementary cumulative distribution functions for total releases are virtually unchanged. Therefore, the corrections to the spallings volume calculation did not impact Waste Isolation Pilot Plant performance assessment calculation results.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
PROJECT FEVER – Fostering Electric Vehicle Expansion in the Rockies
Project FEVER (Fostering Electric Vehicle Expansion in the Rockies) is a part of the Clean Cities Community Readiness and Planning for Plug-in Electric Vehicles and Charging Infrastructure Funding Opportunity funded by the U.S. Department of Energy (DOE) for the state of Colorado. Tasks undertaken in this project include: Electric Vehicle Grid Impact Assessment; Assessment of Electrical Permitting and Inspection for EV/EVSE (electric vehicle/electric vehicle supply equipment); Assessment of Local Ordinances Pertaining to Installation of Publicly Available EVSE;Assessment of Building Codes for EVSE; EV Demand and Energy/Air Quality Impacts Assessment; State and Local Policy Assessment; EV Grid Impact Minimization Efforts; Unification and Streamlining of Electrical Permitting and Inspection for EV/EVSE; Development of BMP for Local EVSE Ordinances; Development of BMP for Building Codes Pertaining to EVSE; Development of Colorado-Specific Assessment for EV/EVSE Energy/Air Quality Impacts; Development of State and Local Policy Best Practices; Create Final EV/EVSE Readiness Plan; Develop Project Marketing and Communications Elements; Plan and Schedule In-person Education and Outreach Opportunities.
33 ADVANCED PROPULSION SYSTEMS↗
Summary Report for ASC L2 Milestone #4782: Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes
This document serves as a detailed companion to the powerpoint slides presented as part of the ASC L2 milestone review for Integrated Codes milestone #4782 titled “Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes”, due on 9/30/2014, and presented for formal program review on 9/12/2014. The program review committee is represented by Mike Zika (A Program Project Lead for Kull), Brian Pudliner (B Program Project Lead for Ares), Scott Futral (DEG Group Lead in LC), and Mike Glass (Sierra Project Lead at Sandia). This document, along with the presentation materials, and a letter of completion signed by the review committee will act as proof of completion for this milestone.
97 MATHEMATICS AND COMPUTING↗
Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation
Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code
Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Initial Assessment of CTF for Time-at-Temperature Applications
The US nuclear industry is interested in improving the economics of their fleet of light-water reactors (LWRs) by uprating US plants. One option being considered is to regain lost margin from overly conservative fuel safety limits. The current limit requires avoidance of critical heat flux (CHF) and prevents further operation of fuel that experiences a dry-out in boiling water reactors (BWRs) or departure from nucleate boiling (DNB) in pressurized water reactors (PWRs); however, it has been shown that temporary, mild dry-out of the fuel does not necessarily increase the risk of fuel failure during its normal anticipated operating life. Such mild dry-out or DNB events may occur during a plant anticipated operational occurrence (AOO), such as a locked rotor in a PWR or a pump trip in a BWR. The time-at-temperature (TAT) approach to regulating fuel operation aims to demonstrate that the fuel rod’s integrity is not challenged during such a mild transient that leads to CHF in which the fuel operates at an elevated temperature for a brief period of time. However, implementing this approach will require extensive fuel material experimental data, as well as supporting modeling and simulation (M&S) predictions, to ensure that the predicted fuel response during AOOs, with all applicable uncertainty considered, will not threaten the safety of the fuel during the transient or the remainder of its anticipated lifecycle. To address this need, a comprehensive effort is being proposed that includes generating cladding material data under TAT conditions, assessment of available code capabilities for TAT conditions, development of new mechanistic models, and demonstration of the M&S capabilities for AOOs of interest. This will require a joint effort between the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Advanced Fuels Campaign (AFC) programs, as well as close collaboration with nuclear industry stakeholders. The outcome of this collaboration will result in development and assessment of capabilities that can be used by the nuclear industry to support qualification of a TAT-based fuel failure criteria safety limit. This report focuses on the thermal hydraulics (T/H) modeling capabilities and summarizes currently available data for validating the T/H subchannel code CTF for TAT conditions, as well as preliminary assessment results of the code. The initial assessment also resulted in implementation of an alternative post-CHF heat transfer package, which has been shown to significantly improve accuracy. This report is not a final assessment and does not consider all available validation data; it is intended that a future assessment will more fully validate the code for this application.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
An Overview of the Monte Carlo Methods, Codes, & Applications Group
This report sketches the work of the Group to deliver first-principle Monte Carlo methods, production quality codes, and radiation transport-based computational and experimental assessments using the codes MCNP and MCATK for such applications as criticality safety, non-proliferation, nuclear energy, nuclear threat reduction and response, radiation detection and measurement, radiation health protection, and stockpile stewardship.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Assessment of Multi-Scale Thermal-Hydraulic Codes and Models for DNB Challenge Problem Applications
The objectives of this milestone work are to assess the improved capabilities of the multi-scale thermal-hydraulic (T/H) codes and models developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) for Departure from Nucleate Boiling (DNB) Challenge Problem (CP), in accordance with the DNB CP implementation plan. The multi-scale models can range from fine mesh Computational Fluid Dynamics (CFD) simulation of flow field surrounding a fuel rod to a full core modeling of a Pressurized Water Reactor (PWR). The assessments are performed based on the CASL subchannel code COBRA-TF (CTF) and the CFD code Hydra-TH. Significant improvements have been achieved on the CASL multi-scale T/H code and modeling capabilities based on the CTF subchannel code and the Hydra-TH CFD code in the past year for DNB CP applications. The improvements are reflected in the new transient modeling for the Reactivity Insertion Accident (RIA) DNB predictions and full assembly modeling using the CTF code and its processor, as well as the rod bundle modeling for single-phase flow and heat transfer simulations using the Hydra-TH code, the pre-processor for mesh generation and the post-processor for data visualization.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
RADTRAN 6 Technical Manual
This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Transport model predictions of 225 Ac production cross sections via energetic p, d and α irradiation of 232 Th targets
Monte Carlo transport codes PHITS and MCNP6 were used to calculate the production cross sections of 225,227 Ac, 227,229 Th, 223,225 Ra, and 229,230,231 Pa via the bombardment of a 232 Th target with energetic protons, deuterons, and α-particles. The incident projectile energies ranged between 10 and 800 MeV/nucleon. When possible, the predicted production cross sections were compared with the available experimental data and other predictions. Here, the degree of the codes’ abilities to match the measured data provides a qualitative assessment of the codes’ abilities to predict data from similar, but unmeasured, projectile/target systems. In addition, a comparison between calculated cross sections and data may provide insight into possible improvements in the physics models employed by those transport codes.
223Ra↗
SU-F-18C-09: Assessment of OSL Dosimeter Technology in the Validation of a Monte Carlo Radiation Transport Code for CT Dosimetry
Purpose: To assess the utility of optically stimulated luminescent (OSL) dosimeter technology in calibrating and validating a Monte Carlo radiation transport code for computed tomography (CT). Methods: Exposure data were taken using both a standard CT 100-mm pencil ionization chamber and a series of 150-mm OSL CT dosimeters. Measurements were made at system isocenter in air as well as in standard 16-cm (head) and 32-cm (body) CTDI phantoms at isocenter and at the 12 o'clock positions. Scans were performed on a Philips Brilliance 64 CT scanner for 100 and 120 kVp at 300 mAs with a nominal beam width of 40 mm. A radiation transport code to simulate the CT scanner conditions was developed using the GEANT4 physics toolkit. The imaging geometry and associated parameters were simulated for each ionization chamber and phantom combination. Simulated absorbed doses were compared to both CTDI{sub 100} values determined from the ion chamber and to CTDI{sub 100} values reported from the OSLs. The dose profiles from each simulation were also compared to the physical OSL dose profiles. Results: CTDI{sub 100} values reported by the ion chamber and OSLs are generally in good agreement (average percent difference of 9%), and provide a suitable way to calibrate doses obtained from simulation to real absorbed doses. Simulated and real CTDI{sub 100} values agree to within 10% or less, and the simulated dose profiles also predict the physical profiles reported by the OSLs. Conclusion: Ionization chambers are generally considered the standard for absolute dose measurements. However, OSL dosimeters may also serve as a useful tool with the significant benefit of also assessing the radiation dose profile. This may offer an advantage to those developing simulations for assessing radiation dosimetry such as verification of spatial dose distribution and beam width.
61 RADIATION PROTECTION AND DOSIMETRY↗
Calibration and validation of NRC fuel performance models
FRAPCON and FRAPTRAN are the steady-state and transient fuel performance codes, respectively, that are used by the U.S. Nuclear Regulatory Commission (NRC) in performing confirmatory calculations of licensee fuel performance codes and safety analysis methods during the review and approval of these codes and methods for light-water reactors. The NRC also uses these codes to perform calculations to direct regulatory research and to assist in the development of new regulations and regulatory guides. FRAPCON and FRAPTRAN are well-validated fuel performance codes that are kept in a constant state of maintenance. FRAPCON and FRAPTRAN have an extensive database of separate effects data used to validate individual material property and fuel performance models and an integral effects database used to assess the overall code prediction within the full range of fuel operations. The key code predictions that are of regulatory interest as they relate to various specified acceptable design limits are: fuel and cladding temperature, rod internal pressure, cladding hoop strain, cladding corrosion and hydriding, and fission gas release. To assess the ability of the code to accurately predict these values, the integral assessment database consists of measurements of fuel centerline temperature, fission gas release, void volume, cladding corrosion, and cladding hoop strain. This paper will discuss the activities that are performed to ensure that FRAPCON and FRAPTRAN represent the state of the industry and will describe the separate effect and integral effect databases that are used to validate the models and are continually being expanded to ensure validation over the full range of fuel operations.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗