Results for “Coding Guidelines” — Page 10

DOE OSTI2015

MELCOR Computer Code Manuals Volume 3: MELCOR Assessment Problems [Draft]

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light-water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC) as a second-generation plant risk assessment tool and the successor to the Source Term Code package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system (RCS), reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; and fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 2.0, released to users in September 2008. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR User's Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package. Volume 3, MELCOR Assessment Problems, presents a portfolio of test and sample problems consisting of both analyses of experiments and of full plant problems. These analyses will be repeated with future releases of MELCOR in order to provide a metric on code predictions as new versions are released.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2017

Guidelines for VCCT-Based Interlaminar Fatigue and Progressive Failure Finite Element Analysis

This document is intended to detail the theoretical basis, equations, references and data that are necessary to enhance the functionality of commercially available Finite Element codes, with the objective of having functionality better suited for the aerospace industry in the area of composite structural analysis. The specific area of focus will be improvements to composite interlaminar fatigue and progressive interlaminar failure. Suggestions are biased towards codes that perform interlaminar Linear Elastic Fracture Mechanics (LEFM) using Virtual Crack Closure Technique (VCCT)-based algorithms [1,2]. All aspects of the science associated with composite interlaminar crack growth are not fully developed and the codes developed to predict this mode of failure must be programmed with sufficient flexibility to accommodate new functional relationships as the science matures.

Deobald, Lyle R.↗

DOE OSTI2022

Release of Contaminants from Burning Liquids and Solids

This report details model development, theory, and a literature review focusing on the emission of contaminants on solid substrates in fires. This is the final report from a 2-year Nuclear Safety Research and Development (NSRD) project. The work represents progress towards a goal of having modeling and simulation capabilities that are sufficiently mature and accurate that they can be utilized in place of physical tests for determining safe handling practices. At present, the guidelines for safety are largely empirically based, derived from a survey of existing datasets. This particular report details the development, verification and calibration of a number of code improvements that have been implemented in the SIERRA suite of codes, and the application of those codes to three different experimental scenarios that have been subject of prior tests. The first scenario involves a contaminated PMMA slab, which is exposed to heat. The modeling involved a novel method for simulating the viscous diffusion of the particles in the slab. The second scenario involved a small pool fire of contaminated combustible liquid mimicking historical tests and finds that the release of contaminants has a high functionality with the height of the liquid in the container. The third scenario involves the burning of a contaminated tray of shredded cellulose. A novel release mechanism was formulated based on predicted progress of the decomposition of the cellulose, and while the model was found to result in release that can be tuned to match the experiments, some modifications to the model are desirable to achieve quantitative accuracy.

54 ENVIRONMENTAL SCIENCES↗

NASA NTRS2008

Assessment of Near-Field Sonic Boom Simulation Tools

A recent study for the Supersonics Project, within the National Aeronautics and Space Administration, has been conducted to assess current in-house capabilities for the prediction of near-field sonic boom. Such capabilities are required to simulate the highly nonlinear flow near an aircraft, wherein a sonic-boom signature is generated. There are many available computational fluid dynamics codes that could be used to provide the near-field flow for a sonic boom calculation. However, such codes have typically been developed for applications involving aerodynamic configuration, for which an efficiently generated computational mesh is usually not optimum for a sonic boom prediction. Preliminary guidelines are suggested to characterize a state-of-the-art sonic boom prediction methodology. The available simulation tools that are best suited to incorporate into that methodology are identified; preliminary test cases are presented in support of the selection. During this phase of process definition and tool selection, parallel research was conducted in an attempt to establish criteria that link the properties of a computational mesh to the accuracy of a sonic boom prediction. Such properties include sufficient grid density near shocks and within the zone of influence, which are achieved by adaptation and mesh refinement strategies. Prediction accuracy is validated by comparison with wind tunnel data.

Casper, J. H.↗

DOE OSTI2024

Challenges in Rotor Aerodynamic Modeling for Non-Uniform Inflow Conditions

Within an international collaboration framework, the accuracy of rotor aerodynamic models used for design load calculations of wind turbines is being assessed. Where the use of high-fidelity computation fluid dynamics (CFD) and mid-fidelity free-vortex wake (FVW) models has become commonplace within the wind energy community, these still fail to meet the requirements in terms of execution time and computational cost needed for design load calculations. The fast but engineering fidelity blade-element/momentum (BEM) method can therefore still be considered the industry workhorse for design load simulations. At the same time, upscaling of wind turbine rotors makes inflow non-uniformities (e.g. shear, veer, turbulence) more important. The objective of this work is to assess model accuracy in non-uniform inflow conditions, which violate several BEM assumptions. Thereto a comparison in turbulent inflow has been executed including a wide variety of codes, focusing on the DanAero field measurements, where a 2.3-MW turbine was equipped with, among other sensors, a pressure measurement apparatus. The results indicate that, although average load patterns are in good agreement, this does not hold for the unsteady loads that drive fatigue damage and aero-elastic stability. A simplified comparison round in vertical shear was initiated to investigate the observed differences in a more controlled manner. A consistent offset in load amplitude was observed between CFD and free-vortex codes on the one hand and BEM-type codes on the other hand. To shed more light on the observations, dedicated efforts are ongoing to pinpoint the cause for these differences, in the end leading to guidelines for an improved BEM implementation.

17 WIND ENERGY↗

DOE OSTI2021

Ducted Assembly Steady-State Heat Transfer Software (DASSH)

The Ducted Assembly Steady-State Heat transfer software (DASSH) is being developed at Argonne National Laboratory to perform a steady-state thermal fluids calculation to determine the coolant flow and temperature distribution for a hexagonal reactor core configuration with ducted assemblies. DASSH is intended to be used in the early stages of the reactor design process when assembly components can be undefined or are undergoing considerable design changes. DASSH provides a rapid assessment of the temperature and flow distribution to allow quick characterization of the system and identification of problem areas. This document is a guide for DASSH users. It summarizes the core capabilities of DASSH and highlights differences between DASSH and SE2-ANL. Instructions for obtaining and installing DASSH are provided along with some guidelines and recommendations for newer Python users. It covers the structure of the input file, provides directions for running DASSH, describes the various output files produced by the code, and provides instructions for visualizing data. Examples demonstrating various DASSH options are included. DASSH is under active development, so it is anticipated that this guide will grow and evolve as the code is updated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1996

Bounds on Block Error Probability for Multilevel Concatenated Codes

Maximum likelihood decoding of long block codes is not feasable due to large complexity. Some classes of codes are shown to be decomposable into multilevel concatenated codes (MLCC). For these codes, multistage decoding provides good trade-off between performance and complexity. In this paper, we derive an upper bound on the probability of block error for MLCC. We use this bound to evaluate difference in performance for different decompositions of some codes. Examples given show that a significant reduction in complexity can be achieved when increasing number of stages of decoding. Resulting performance degradation varies for different decompositions. A guideline is given for finding good m-level decompositions.

Lin, Shu↗

NASA NTRS1973

A digital communications system for manned spaceflight applications.

A highly efficient, all-digital communications signal design employing convolutional coding and PN spectrum spreading is described for two-way transmission of voice and data between a manned spacecraft and ground. Variable-slope delta modulation is selected for analog/digital conversion of the voice signal, and a convolutional decoder utilizing the Viterbi decoding algorithm is selected for use at each receiving terminal. A PN spread spectrum technique is implemented to protect against multipath effects and to reduce the energy density (per unit bandwidth) impinging on the earth's surface to a value within the guidelines adopted by international agreement. Performance predictions are presented for transmission via a TDRS (tracking and data relay satellite) system and for direct transmission between the spacecraft and earth. Hardware estimates are provided for a flight-qualified communications system employing the coded digital signal design.

Batson, B. H.↗

DOE OSTI2015

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2016

Reactor Safety for Beyond Design Basis Events

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant's current design basis. The DOE has played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R and D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating as well as better characterize the safety performance of future U.S. nuclear power plants. In pursuing this area of R and D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary end user of the results from this DOE-sponsored work. The purpose of the Reactor Safety Technology R and D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet. The RST R and D activities have evolved from an initial coordinated international effort to assist in the analysis of the Fukushima accident progression and accident response into the following three areas of current work. Fukushima Forensics and Examination Plans: This R and D is focused on providing insights into the actual severe accident progression at Fukushima through planning and interpretation of visual examinations and data collection of in-situ conditions of the damaged units as well as collection and analysis of samples from the damaged reactors. Severe Accident Analyses: This R and D is focused on analyses using existing computer models and their ability to provide information and insights into severe accident progression that aid in the development of severe accident management guidelines (SAMG) and/or training operators on these SAMGs; an auxiliary benefit can be informing improvements in these models. Accident Tolerant Components: This R and D work is focused on analysis or experimental efforts for hardware-related issues, including systems, structures and components with the potential to prevent core degradation or mitigate the effects of beyond-design basis events. This presentation will review current and future R and D activities in this reactor safety area. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1994

C style guide

This document discusses recommended practices and style for programmers using the C language in the Flight Dynamics Division environment. Guidelines are based on generally recommended software engineering techniques, industry resources, and local convention. The Guide offers preferred solutions to common C programming issues and illustrates through examples of C Code.

Doland, Jerry↗

NASA NTRS2011

Preliminary Computational Analysis of the (HIRENASD) Configuration in Preparation for the Aeroelastic Prediction Workshop

This paper presents preliminary computational aeroelastic analysis results generated in preparation for the first Aeroelastic Prediction Workshop (AePW). These results were produced using FUN3D software developed at NASA Langley and are compared against the experimental data generated during the HIgh REynolds Number Aero- Structural Dynamics (HIRENASD) Project. The HIRENASD wind-tunnel model was tested in the European Transonic Windtunnel in 2006 by Aachen University0s Department of Mechanics with funding from the German Research Foundation. The computational effort discussed here was performed (1) to obtain a preliminary assessment of the ability of the FUN3D code to accurately compute physical quantities experimentally measured on the HIRENASD model and (2) to translate the lessons learned from the FUN3D analysis of HIRENASD into a set of initial guidelines for the first AePW, which includes test cases for the HIRENASD model and its experimental data set. This paper compares the computational and experimental results obtained at Mach 0.8 for a Reynolds number of 7 million based on chord, corresponding to the HIRENASD test conditions No. 132 and No. 159. Aerodynamic loads and static aeroelastic displacements are compared at two levels of the grid resolution. Harmonic perturbation numerical results are compared with the experimental data using the magnitude and phase relationship between pressure coefficients and displacement. A dynamic aeroelastic numerical calculation is presented at one wind-tunnel condition in the form of the time history of the generalized displacements. Additional FUN3D validation results are also presented for the AGARD 445.6 wing data set. This wing was tested in the Transonic Dynamics Tunnel and is commonly used in the preliminary benchmarking of computational aeroelastic software.

Chwalowski, Pawel↗

DOE OSTI2016

SU-F-I-02: Comparative Analysis and Constancy Check of Image Quality Parameters for Three Linear Accelerators Per TG 142 Protocol

Purpose: To compare image quality parameters and assessing the image stability of three different linear accelerators (linac) for 2D and 3D imaging modalities: planar kV, MV images and cone-beam CT (CBCT). Methods: QCkV1, QC-3 and Cathpan-600 phantoms were utilized to acquire kV, MV and CBCT images respectively on monthly basis per TG142 QA protocol for over 2 years on 21Ex, NovalisTx and TrueBeam linacs. DICOM images were analyzed with the help of QA analysis software: PIPsPro from Standard Imaging. For planar kV and MV images, planar spatial resolution, contrast to noise ratio (CNR) and noise; for CBCT, HU values were collected and analyzed. Results: Two years of monthly QA measurements were analyzed for the planar and CBCT images. Values were normalized to the mean and the standard deviations (STD) are presented. For the kV planar radiographic images the STD of spatial resolution for f30, f40, f50, CNR and noise for 21Ex are 0.006, 0.011, 0.013, 0.046, 0.026; Novalis-Tx are 0.009, 0.016, 0.016, 0.067, 0.053 ; TrueBeam are 0.007, 0.005, 0.009, 0.017, 0.016 respectively. For the MV planar radiographic images, the STD of spatial resolution for f30, f40, f50, CNR and noise for 21Ex are 0.009, 0.010, 0.008, 0.023, 0.023; for Novalix-Tx are 0.012, 0.010, 0.008, 0.029, 0.023 and for TrueBeam are 0.010, 0.010, 0.007, 0.022, 0.022 respectively. For the CBCT images, HU constancies of Air, Polystyrene, Teflon, PMP, LDPE and Delrin for 21Ex are 0.014, 0.070, 0.031, 0.053, 0.076, 0.087; for Novalis Tx are 0.019, 0.047, 0.035, 0.059, 0.077, 0.087 and for TrueBeam are 0.011, 0.044, 0.025, 0.044, 0.056, 0.020 respectively. Conclusion: These Imaging QA results demonstrated that the TrueBeam, performed better in terms of image quality stability for both kV planer and CBCT images as well as EPID MV images, however other two linacs were also satisfied TG142 guidelines.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2016

Energy Savings Modeling and Inspection Guidelines for Commercial Building Federal Tax Deductions for Buildings in 2016 and Later

This document provides guidance for modeling and inspecting energy-efficient property in commercial buildings for certification of the energy and power cost savings related to Section 179D of the Internal Revenue Code (IRC) enacted in Section 1331 of the 2005 Energy Policy Act (EPAct) of 2005, noted in Internal Revenue Service (IRS) Notices 2006-52 (IRS 2006), 2008-40 (IRS 2008) and 2012-26 (IRS 2012), and updated by the Protecting Americans from Tax Hikes (PATH) Act of 2015. Specifically, Section 179D provides federal tax deductions for energy-efficient property related to a commercial building's envelope; interior lighting; heating, ventilating, and air conditioning (HVAC); and service hot water (SHW) systems. This document applies to buildings placed in service on or after January 1, 2016.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2015

The 2014 Sandia Nonlinear Mechanics and Dynamics Summer Research Institute

A collaborative research institute was organized and held at Sandia Albuquerque for a period of six weeks. This research institute brought together researchers from around the world to work collaboratively on a set of research projects. These research projects included: developing experimental guidelines for studying variability and repeatability of nonlinear structures; decoupling aleatoric and epistemic uncertainty in measurements to improve dynamic predictions; a numerical round robin to assess the performance of five different numerical codes for modeling systems with strong nonlinearities; and an assessment of experimentally derived and numerically derived reduced order models. In addition to the technical collaborations, the institute also included a series of seminars given by both Sandians and external experts, as well as a series of tours and field trips to local places of scientific and engineering importance. This report details both the technical research and the programmatic organization of the 2014 Sandia Nonlinear Mechanics and Dynamics Summer Research Institute.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Pellet cladding mechanical interaction as a potential failure mechanism during a control rod drop accident in a boiling water reactor

Boiling water reactors (BWRs) represent approximately one-third of the operating fleet in the United States, contributing significantly towards the global effort in reducing carbon emissions. Even though most of the operating fleet has been in operation for quite some time, continued advancements in new nuclear fuel (such as accident tolerant fuel) or operating regimes (such as power up-rates and higher burnup operation) necessitates similar advancements in modeling and simulation capabilities. Bison, a component of the Virtual Environment for Reactor Applications (VERA), is a high-fidelity fuel performance code able to explore the fuel performance of a wide variety of fuel types in one-, two-, and three-dimensions. Until recently, the code had not been used for analyses of BWRs. Modeling capabilities have been added for Gd-bearing UO{sub 2} and pure zirconium liners. New models have been added based upon the U.S. Nuclear Regulatory Commission (NRC) guide-lines for hydrogen pickup in Zircaloy-2 claddings and pellet-clad mechanical interaction (PCMI) failure during a reactivity insertion accident (RIA), known as a rod drop accident (CRDA) in BWRs. Implementing and/or improving Bison modeling capabilities extended its analytical reach to areas beyond its original intended purpose. This paper demonstrates one of these capabilities as a proof of concept. Recently developed models enable Bison to provide an alternate approach for cladding integrity determination in CRDA evaluations, which currently use bounding conservative estimates. As part of this demonstration, NRC guidance on hydrogen-pickup and PCMI failure models were utilized in this research. Even though more research is needed in establishing right inputs and process in this area, this paper demonstrates Bison's ability to determine cladding integrity in a CRDA evaluation. This first of a kind demonstration is a proof of concept in this area, which could potentially be extended to a number of other areas where a more accurate cladding integrity determination would be needed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗