It is prerequisite for the establishment of nuclear power plant safety and for the maximization of operation efficiency to devlope the accident analysis computer code packages which can predict the results of postulated accidents and evaluate the performa...
National Technical Information Service (NTIS)
The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The VICTORIA code is designed to model the in-vessel phase of a severe reactor accident. The functions of VICTORIA in MELPROG are fission product release from fuel,...
In the analysis of accidents a collision between aircraft is treated as one accident in the overall total. However, a complete analysis and coding is made on each aircraft involved in a collision. This produces two aircraft accident records, one for each ...
... Title : Uncertainty and Sensitivity Analysis of a Fire-Induced Accident Scenario Involving Binary Variables and Mechanistic Codes. ...
DTIC Science & Technology
The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...
The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...
A computer code called TRAC is being developed by the Los Alamos Scientific Laboratory for analysis of loss-of-coolant accidents (LOCA's) and other transients in light water reactors. This code differs from existing codes and other codes under development...
This report describes the analytical results of operational transient and accidents for the safety assessment of upgraded JRR-3. The following six items of transients and accidents have been selected and analyzed for the assessment; 1. decrease of primary...
The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reac...
The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accid...
The accident at Three Mile Island Unit 2 (M-2) provides an opportunity to benchmark severe accident analysis methods against full-scale, integrated facility data. In collaboration with the US Department of Energy (DOE), the OECD Nuclear Energy Agency established a joint task group to analyze various periods of the ...
Energy Citations Database
Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. ...
Sandia has made considerable progress in the past year on the MELCOR code for integrated severe nuclear reactor accident analysis. Actinities for the past year are presented.
DOE Information Bridge
The MARCH and HECTR computer codes are used in this study to examine hydrogen production, transport, and combustion in an ice-condenser containment for a number of hypothesized severe accidents. Both degraded-core and core-meltdown accidents are treated. ...
The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and ...
The Department of Energy has developed new capabilities to predict how an accident event would spread through a facility. These capabilities arose as a part of the development of a future generation accident code - CONACS. They can be used to predict the ...
One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat t...
The SAS4A code system is a new tool for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failures of the subassembly walls. The objective in the development of SAS4A is to provide improved analytical mod...
During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident an...
This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and...
This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in ...
Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes ...
This paper describes methods, in the form of a handbook and five computer codes, that can be used for air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ...
The rod ejection accident (REA) of the Beznau (KKB-2) nuclear power plant was investigated. The REA analysis was performed using the RETRAN-02 computer code. Four basic cases were investigated for the cycle 16 conditions. At the beginning-of-life (BOL) th...
A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting f...
The RELAP5/MOD1 computer code was used in the analysis of a loss coolant accident (LOCA), postulated for Angra-2. The power plant was simulated through a division in control volumes and junctions suitable for a small break accident calculation. Initially ...
A number of computer codes have been written to attempt to predict the consequences of hypothetical accidents. These codes all have one thing in common: they produce reams of paper containing vast quantities of numbers. In an effort to make these numbers ...
The VICTORIA code is designed to be a module of the MELPROG severe accident analysis code. The functions of VICTORIA in MELPROG are fission product release from fuel, vapor-phase and aerosol physics, and chemical interactions involving fuel, structures, c...
The TRAC (Transient Reactor Analysis Code) program is an advanced computer code which is designed to model postulated accidents in light water reactors. The paper discusses some of the methodology employed in modeling large internal flow systems. Emphasis is placed on the numerical and modeling problems inherent in ...
The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, ...
The Transient Reactor Analysis Code (TRAC) is being developed to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in light water reactors. The TRAC-PD2 program provides this analysis capability for pressuriz...
... Descriptors : (*AUSTRALIA, ACCIDENTS), (*ACCIDENTS, STATISTICAL ANALYSIS), MOTOR VEHICLE ACCIDENTS, INDUSTRIAL MEDICINE ...
MARCH 2 calculations for the TMLB' accident (a transient with loss of all ac power) at Surry provide a test bed for addressing problems with computer models in an analysis situation typical of those expected in applications of the MELCOR code system for severe accident analysis. This ...
The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major ...
INTRODUCTION: The rate for work related accidents in the Spanish mining sector is notably higher than in other countries such as the United States. It produces a very negative impact on the mining industry. This paper is the report of a study on serious and fatal accidents in Spanish mining from 1982-2006. It is based on the reports of 212 ...
PubMed
Accurate predictions for the normal, off-normal or accident conditions are required for the safety analysis to evaluate the plant safety. Safety analysis computer codes such as IANUS, DEMO have been specifically developed for the Fast Flux Test Facility, ...
This report documents the results of an assessment to determine the reactor and containment hardware, systems, and phenomena which must be modeled in realistic boiling water reactor severe accident analysis computer codes. The scope of the assessment is l...
This study is an evaluation of a pressurized-water reactor (PWR) accident as defined by WASH 1400 for the proposed nuclear reactor site at Cementon, N. Y. Using an extension of the Environmental Protection Agency's AIREM computer code, the following were ...
This volume is the second of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. Thes...
This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These...
The Three Mile Island nuclear plant (TMI-2) was modeled using the Transient Reactor Analysis Code (TRAC-PIA) and a preliminary calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. The purpose of thi...
Sandia National Laboratories is participating in several NRC-sponsored programs to study severe accident phenomenology. Part of that effort involves the combined use of the computer codes MARCH and HECTR to examine hydrogen behavior during severe accident...
The report documents the procedures and results of an analysis of preventable accidents involving motor carriers. The process used to assign 'fault' to drivers, vehicles, and owners is explained. Also, data sources, coding conventions and analytical techn...
The CRAC2 computer code has been adapted to the calculation requirements of Draft/Final Environmental Impact Statement (DES/FES) casework analysis for the Nuclear Regulatory Commission. CRAC2 is a revised version of the CRAC (Calculation of Reactor Accide...
CORCON is a computer code for modelling the interactions between molten core materials and concrete, such as might occur following a core meltdown accident in a Light Water Reactor. It may also be applied to experiments which simulate such accident condit...
The purpose of the work described in this paper was to evaluate the adequacy of criticality accident alarm system (CAAS) coverage of several buildings at the Portsmouth Gaseous Diffusion Plant (PORTS) located in Piketon, Ohio. An analysis was performed in which the General Monte Carlo N-Particle Transport Code (MCNP) was used to model ...
The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the saf...
codes Q and M. The Challenger accident. (1986) impacted the development of this technology in several ways. First, Professor. Richard. Feynman, ...
NASA Website
is analyzing response of paratranist bus structure to rollover accident using LS-DYNA code. Click for Details Transims visualization Transims visualization TRANSIMS...
Science.gov Websites
... The computer codes RELAP-3 for reactor system blowdown and THETAl-B for fuel element heatup were used to perform these calculations at a ...
The paper gives an analysis of the current state of the RBMK safety evaluation in accidents initiated by partial ruptures of the delivery part of the circulating loop. It appears from this analysis that applicability and uncertainty of the international code RELAP for RBMK safety analysis could ...
The Korea Electric Power Research Institute had decided to develop the new safety analysis code system for the Optimized Power Reactor 1000 (OPR1000) in Korea by the fund of the Ministry of Commerce, Industry and Energy. In this paper, some results of the Advanced Power Reactor 1400(APR1400) using the RETRAN code for some reactivity ...
The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe ...
have been examined using the codes RELAP, TRAC, RAMONA, and SMABRB. Severe accidents have been-of-coolant accident analysis. Pour differ- ent codes have been studied during 1982. The American code RELAP5 has been and to the RELAP5 calcu- lations of Finland and Sweden. The ...
E-print Network
For many years, study of whole-core transients and accidents in fast reactors has been a major element of safety research and development programs in the US. Numerous models and computer codes have been developed and validated for use in these studies. Historically, emphasis has been placed on describing the core disruptive accident, ...
Accident models play a critical role in accident investigation and analysis. Most traditional models
NASA Technical Reports Server (NTRS)
This paper describes a multilaboratory research program that is directed toward addressing many questions that analysts face when performing air cleaning accident consequence assessments. The program involves developing analytical tools and supportive experimental data that will be useful in making more realistic assessments of accident source terms within ...
This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to ...
An international analysis exercise has been organized to evaluate the ability of nuclear reactor severe accident computer codes to predict the TMI-2 accident sequence and core damage progression during the first 300 minutes of the accident. A required bou...
This chapter discusses the development of current state-of-the-art computational methods used in the United States for analysis of core meltdown accidents (Hypothetical Core Disruptive Accidents) in LMFBRs. The emphasis is on the phenomenological basis and numerical methods of the codes SAS, VENUS-II, SIMMER, and ...
The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident compute...
This paper provides an overview of the current activities within the international consortium participating in the Three Mile Island Unit 2 (TMI-2) analysis exercise, which is part of the Nuclear Energy Agency/Department of Energy Joint Task Group (JTG) program on TMI-2, formed to utilize the TMI-2 accident as a benchmark for ...
This report presents the comparisons between calculated and measured fuel rod behavior and the analysis of stress for preirradiated LWR type fuel rods during reactivity initiated accident (RIA) conditions. For the calculations, FPRETAIN computer code whic...
The SIMMER Code, developed at Los Alamos Scientific Laboratory, analyzes the various phases of a CDA (Core disassembly accident) in an LMFBR. The qualification testing program, comprised of two stages, is discussed. The first stage is developmental verifi...
An update is given on very recent developments involving CONTAIN, the USNRC's principal mechanistic code for severe accident containment analysis. First, the features are outlined in two major new releases of CONTAIN. Revision 1.06 was released in Februar...
Technical progress during this quarter has been concerned primarily with continued development of the TRAP-MELT computer code, specification of conditions and code input for meltdown accidents, and initiation of efforts to provide functional design specif...
The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the ...
Regardless low probability of occurrence the severe accident phenomena are investigated for all types of nuclear reactors in the world because the consequences of such accident could be catastrophic. Most of research is performed for the prevailing vessel-type light water reactors like PWRs and BWRs. Less research is performed for the channel-type reactors ...
To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze ...
The MELCOR code, developed by Sandia National Laboratories, is a fully integrated, relatively fast-running code that models the progression of severe accidents in commercial light water nuclear power plants (NPPs).A specific station blackout (SBO) accident for Kuosheng (BWR-6) NPP is simulated using the MELCOR ...
An assessment of the radiological consequences that would result for the population of Greece from postulated major nuclear accidents in the Kozloduy nuclear power station in Bulgaria is performed. Kozloduy lies at a distance of 225 km from the northern borders of Greece and contains six reactors, all of the Russian WWER type. The postulated accidents that ...
This study evaluates the effect that the timing of vessel depressurization has in the progression of a short-term station blackout accident in a boiling water reactor (BWR)-4. This study was performed with the MELCOR (version 1.8.1) severe-accident code. A similar study was previously completed at Oak Ridge National Laboratory (ORNL) ...
... Accession Number : AD0620258. Title : FURTHER ATTEMPTS AT CODING AIRCRAFT ACCIDENTS. Descriptive Note : Research rept.,. ...
... AD0076065. Title : CODED FLEET ACCIDENTS OF RECENT GRADUATES. Corporate Author : TULANE UNIV NEW ORLEANS LA. ...
SAMPSON is the integral code for severe accident analysis in detail with modular structure, developed in the IMPACT project. Each module can run independently and communication with multiple analysis modules supervised by the analysis control module makes an integral ...
NASA Astrophysics Data System (ADS)
Response surface techniques have been developed for obtaining probability distributions of the consequences of postulated nuclear reactor accidents. The probabilistic response surface methodology reported includes new knot-point selection schemes and response surface functions, functional transformations of both parameters and consequence variables, smooth synthesis of ...
The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for ...
The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and health effects to the public from postulated plutonium releases and from ...
This document describes the procedure used by UNI to develop and use the RELAP4 code for analysis of the N-Reactor system during a LOCA. The code modifications required are described as are the responses to the 10CFR50 App. K requirements. Each of the models used in the analysis is described and discussed. This ...
The Modular Accident Analysis Program (MAAP) simulates LWR system response to a severe core accident. Overall, calculations performed with the PWR version of MAAP have compared well with a wide variety of other data. These results have proven MAAP an acceptable tool to support individual plant examinations and ...
The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) wa...
Within the framework of a danger analysis which serves as a decision-finding basis for emergency planning, rarely expected severe accidents, which are not covered by design basis accidents, have to be considered at the Rossendorf research site with regard...
For a long time in the Russian Federation the computer code for analysis of severe accidents is being developed. The main peculiarity of this code from the known computer codes for analysis of severe accidents at NPP such as MELCOR and ASTEC, is a ...
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted ...
HTGR safety evaluation included studies on fission product release; materials, chemistry, and instrumentation; structural evaluation; and analytical safety evaluation. LMFBR safety evaluation included studies on accident sequences, technical coordination of structural integrity, and SSC code development and validation. LWR safety studies included ...
The report describes the two-dimensional, Pressurized Water Reactor (PWR) version of the MELPROG computer code, MELPROG/PWR-MOD1. The purpose of MELPROG is to provide for an accident sequence a description of (1) the state of the reactor core and surround...
A two-dimensional code FRETA-B was developed to analyze the behaviors of LWR fuel rod bundles under accident conditions. It calculates fuel temperatures, oxidation of claddings, plenum gas pressure and flow, and deformations of fuel rods. Two-dimensional ...
The RELAP4 computer code is a very useful tool for nuclear safety analysis. It is used principally in the analysis of the hypothetical loss-of- coolant accident but is also used in several other applications. A description is given of the basic fluid model; the improvements over its predecessor, RELAP3; ...
Overall, SAS4A appears to do a good job for simulating the OPERA-15 experiment. For most of the experiment parameters, the code calculations compare quite well with the experimental data. The lack of a multi-dimensional voiding model has the effect of extending the flow coastdown time until voiding starts; otherwise, the code simulates the ...
Applications of an Eulerian code to predict the response of LMFBR containment and primary piping systems to hypothetical core disruptive accidents (HCDA), and to analyze sodium spillage problems, are described. The computer code is an expanded version of the ICECO code. Sample problems are presented for ...
The significant features of SIMMER-II, a disrupted-core analysis code, are described. The code has the capabalities to begin space-time neutronics calculations from nonstationary reactor states, to track the intermixing of fuel of different enrichments, and to model the complicated heat- and mass-transfer processes that occur in the ...
Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters whi...
The purpose of this paper is to analyse the significance of the human element in accidents involving Greek-flagged ships, during 19931% of all accidents were attributed to the human element, whereas 754% of the onboard human-induced accidents were linked to errors and violations of the ship's master. Furthermore, since the timeframe ...
The MELRPI computer code has been developed by Rensselaer Polytechnic Institute under the sponsorship of Oak Ridge National Laboratory (ORNL) and, more recently, the Empire State Electrical Energy Research Corporation (ESEERCO). The code was developed especially for severe accident analyses concerning BWRs and is not applicable to ...
The MELCOR computer code is being used as the point of departure to develop an integrated severe accident analysis computer code for the heavy-water moderated U-Al fuel reactors. The resulting computer code (MELCOR/SR) provides a practical and comprehensive analytical tool for evaluating severe ...
The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into...
A modular computational system known as the Water Reactor Analysis Package (WRAP) is being developed at the Savannah River Laboratory for analysis of loss of coolant accidents (LOCA's) and other transients in water reactor systems. At this time, WRAP is e...
Techniques for probabilistic response-surface analysis have been developed to obtain the probability distributions of the consequences of postulated nuclear-reactor accidents. The uncertainties of the consequences are caused by the variability of the syst...
The safety issues associated with the transition (core disruption) phase of hypothetical LMFBR accidents are identified and discussed. This discussion is intended to serve as a basis for defining future experimental needs to support core disruption analysis, and for defining analytical code developments required to support the ...
A great deal of effort has been devoted in the last five years by the reactor safety analysis community to the development of Eulerian computer codes for the combined thermodynamic and hydrodynamic analysis of accident progression. As hydrodynamic calcula...
The consequences of an earthquake-induced fire involving a tritium leak were analyzed using the GASFLOW computer code. Modeling features required by the analysis include ventilation boundary conditions, flow of a gas mixture in an enclosure containing obs...
The consequences of an earthquake-induced fire involving a tritium leak were analyzed using the GASFLOW computer code. Modeling features required by the analysis include ventilation boundary conditions, flow of a gas mixture in an enclosure containing obstacles, thermally induced buoyancy, and combustion phenomena.
The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the ...
The in-depth reviews of risk assessments performed specifically for the Limerick and Shoreham plants have identified accident sequences that are important contributors to core damage frequency (CDF). These plants are boiling water reactors (BWRs) with Mark-II containment designs. Plant features and operator actions that have been found to be important for either preventing or ...
Response-surface techniques have been developed for obtaining probability distributions of the consequences of postulated nuclear reactor accidents. In these techniques, probability distributions are assigned to the system and model parameters of the accident analysis. A limited number of parameter values (called knot points) are ...
The TRAC-BD1 (transient reactor analysis code) provides a consistent and unified analysis capability of the entire loss-of-coolant accident (LOCA) sequence, beginning with the blowdown phase, through heatup, reflood with quenching, and finally the refill ...
The radiological analyses of extreme hypothetical accidents were performed almost wholly by computer techniques. Major analytical codes used were RIBD, CACECO, SPRAY, HAA, and COMRADEX. This paper describes the analyses, the modeling techniques, and transition programs that edited data output from one code into a form suitable for ...
The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the ...
The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and...
The TRAC code (Transient Reactor Analysis Code), developed in the Los Alamos National Laboratory, is used to accident analysis in light water reactor. The TRAC-PD2 version, used in this paper, has a refined dynamic flow model for two fluids, which is base...
A system utilising radiation transport codes has been developed to derive accurate dose distributions in a human body for radiological accidents. A suitable model is quite essential for a numerical analysis. Therefore, two tools were developed to setup a 'problem-dependent' input file, defining a radiation source and an exposed person ...
This report identifies the results of a program aimed at developing a computer code for use in the analysis of the behavior of iodine during steam generator tube rupture (SGTR) accidents in pressurized water reactors (PWR's). The program was directed towards the identification of the several processes that play a role in the ...
Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, ...
