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 ...

This paper is concerned with the application of severe-accident codes to develop improved accident mitigation strategies for nuclear power plants employing boiling water reactors (BWRs). Specifically, the results of analyses using the APRIL.MOD3X code are presented for typical station blackout ...

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...

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 ventilation systems in the ...

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 accident management. This ...

This paper concerns applicability of drywell cooler (DWC) heat removal under severe accident condition in BWR plants. Newly developed heat removal models based on DWC heat removal experiments were built into the MAAP3 code. And then, two types of Japanese BWR were selected to evaluate DWC heat removal performance under typical severe ...

VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from t...

The WECHSL code, developed at Kernforschungszentrum Karlsruhe, West-Germany, is used for core melt accidents in nuclear power plants. The first calculations, considering silicate and limestone/common sand concretes of different compositions, analyze the i...

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...

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) ...

A calculational procedure for the Station Blackout Severe Accident Sequence at Browns Ferry Unit One has been repeated with plant-specific application to one of the Peach Bottom Units. The only changes required in code input are with regard to the primary continment concrete, the existence of sprays in the secondary containment, and ...

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 paper describes a ...

... 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. ...

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 ...

The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, ...

The THERMIX-KONVEK code was used to model the steady state and dynamic thermal behavior of the U.S. pebble bed modular HTR concept and trial calculations for accident conditions were performed. Results of these trial calculations are compared with other p...

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 insertion ...

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 ...

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 transition phase. ...

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 ...

This paper reports on the UFOMOD program system which is an advanced probabilistic accident consequence assessment (ACA) code. Its structure and modeling are based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study Phase A, the results of scientific ...

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 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 not be determined up to the present. At ...

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 ...

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 provide realistic ...

Over the past year, the focus of RELAP5 use at the Savannah River Site has been on code applications to reactor accidents having a direct bearing on setting power limits, with a lesser emphasis on code development. In the applications task, RELAP5/MOD2.5 has been used to predict the ...

Over the past year, the focus of RELAP5 use at the Savannah River Site has been on code applications to reactor accidents having a direct bearing on setting power limits, with a lesser emphasis on code development. In the applications task, RELAP5/MOD2.5 has been used to predict 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, and fission gas release ...

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; and a further ...

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 ...

This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER ...

Dose calculations were performed using the MELCOR Accident Consequence Code System (MACCS) to support safety analyses for the Los Alamos Neutron Science Center (LANSCE) facility. The LANSCE facility is operated and maintained at Los Alamos National Labora...

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...

This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR em...

The acquisition of the ALMOD computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which loss of coolant doesn't occur. The implementation of the German computer code ALMOD and its application ...

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...

... They are Comma codes, Shift codes, B codes, Huffman codes, Conditional, Coded Arithmetic codes, and Two dimensional codes. ...

The objective of the work reported in this paper was to simulate and analyze the in-vessel phenomena of core meltdown and melt progression during hypothetical station blackout accidents in Swedish boiling water reactors (BWRs). The analysis has been performed using the APRIL.MOD3X severe-accident code, with the initial events (i.e., ...

The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated ...

The THERMIX-KONVEK code was used to model the steady state and dynamic thermal behavior of the US pebble bed modular HTR concept and trial calculations for accident conditions were performed. Results of these trial calculations are compared with other predictions by ORNL and by industrial proponents. The basic equations, assumptions, and calculational ...

RETRAN represents a computer code approach for analyzing the thermal hydraulic responses of nuclear steam supply systems to hypothetical loss of coolant-accidents (LOCAs) and operational transients. In contrast to the conservative approach, RETRAN provides best estimate solutions to hypothetical LOCAs and operational transients. RETRAN is a computer ...

The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because ...

The COMRADEX Code is discussed briefly and instructions are provided for the use of the code. The subject code was developed for calculating doses from hypothetical power reactor accidents. It permits the user to analyze four successive levels of containment with time-varying leak rates. Filtration, cleanup, ...

This report is a summary of the material transport modeling procedures developed to support a family of accident analysis computer codes. The material transport modeling areas include transport initiation, convection, interaction, depletion, and filtration. Except for material interaction, these areas are developed in modular form in three Los Alamos ...

GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering ...

The ORNL safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks ...

The integrated SCDAP/RELAP5 code is being developed by the Office of Nuclear Regulatory Research of the USNRC for the purpose of calculating the core damage and fission product transport within the RCS during LWR severe accidents events. Activities since October 1985 were concentrated in three areas. The first area, code and model ...

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...

Work on postulated severe accident sequence code development and application continued both for the Fort St. Vrain and 2240-MW(t) lead plant designs. Initial experiments on high-temperature gas-cooled reactor (HTGR) fission-product release and transport w...

A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer sy...

BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under ...

With respect to the application of the accident consequence model of the German Risk Study (GRS) for light water reactors to risk assessments of other reactor types (high temperature reactor HTR-1160, fast breeder reactor SNR-300), the improved version UF...

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,...

This paper describes a personal computer-based program (GENSOR) that utilizes a simplified time-dependent approach for predicting the radionuclide releases during postulated reactor accidents. This interactive computer program allows the user to generate simplified source terms based on those severe accident attributes that most influence radionuclide ...

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 for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term ...

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 applications to consider problem/model discontinuities in a general fashion, ...

The damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out on a Three Mile Island Unit 2 configuration using the computer code ...

... ACCIDENTS, *RAILROADS, *SAFETY, RAILROADS, DECELERATION, WOUNDS AND INJURIES, EJECTION, RAILROAD CARS, AVIATION ...

in aircraft or rail accidents, the sheer number of road accidents occurring in a ... their total costs usually outstrip those from accidents in other modes. ...

We introduce a new coupled neutronics/thermal hydraulics code system for analyzing transients of nuclear power plants and research reactors, based on a neutron transport theory approach. For the neutron kinetics, we have developed the code DORT-TD, a time-dependent extension of the well-known discrete ordinates code DORT. DORT-TD uses ...

This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can ...

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break ...

The Commission of the European Communities and the Nuclear Energy Agency of the OECD have organized an international exercise to compare the predictions of accident consequence assessment codes, and to identify those features of the models which lead to differences in the predicted results. Alongside this, a further exercise was undertaken in which the ...

The Commission of the European Communities, within the framework of radiation protection research program, initiated a project entitled Methods for Assessing the Radiological Impact of Accidents (MARIA). This project was continued and enlarged within the 1985-1989 research program. The main objectives of this paper are to develop a new probabilistic ...

THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of...

Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initi...

New features have been added to TITAN, which is an advanced 3-D thermohydraulic and neutronic coupled code. The new features include a high order feedback model, a xenon model and a capability to handle nonreactangular (zig-zag) geometry. A central rod ej...

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 document is the primary reference for the ...

ECART can simulate the thermal-hydraulic behavior of LWR and GCR plants under severe accident conditions together with the transport of radio-toxic substances. This tool is still under improvement and assessment for new applications in non-nuclear risk studies, new advanced and fusion reactors. As regards accidents with fires within ...

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many ...

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many ...

As a result of the need for a more accurate computational methodology, the Los Alamos developed Monte Carlo code MCNP is used to show the implementation of a more advanced and accurate methodology in criticality accident detector analysis. This paper will detail the application of MCNP for the analysis of the areas of coverage of a ...

The FRAP-T6 code was developed at the Idaho National Engineering Laboratory (INEL) for the purpose of calculating the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to severe hypothetical loss-of-coolant accidents. An important application of the FRAP-T6 ...

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. ...

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. ...

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 MARCH code, written at Battelle's Columbus Laboratories for the U.S. Nuclear Regulatory Commission, describes the response of LWR systems to accidents which can result in core meltdown. The calculations are performed from the start of the accident thr...

An exploratory sensitivity study with the MELCOR Accident Consequence Code System (MACCS) is presented. This study was performed to provide (1) an indication of the possible impact of consequence modeling uncertainties on the results of an integrated prob...

The initial activities of a Department of Energy (DOE) Safety Analysis Software Group to establish a Safety Analysis Toolbox of computer models are discussed. The toolbox shall be a DOE Complex repository of verified and validated computer models that are configuration-controlled and made available for specific accident analysis applications. The toolbox ...

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 ...

The RELAP4/MOD6 transient analysis code is the most recently released of a set of computer programs designed for calculation of the thermal-hydraulic behavior of a light water reactor (LWR) during the transient phases of a postulated loss-of-coolant accident. Earlier versions of RELAP4 primarily had capability for analysis of blowdown and refill phenomena. ...

The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both ...

Multiphase flow frequently occurs in a progression of accidents of nuclear reactor severe core damage. The CHAMPAGNE code has been developed to analyze thermohydraulic behavior of multiphase and multicomponent fluid, which requires for its characterization more than one set of velocities, temperatures, masses per unit volume, and so forth at each location ...

Space-time neutronic behavior of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated loss-of-coolant accidents. The approach involves the efficient use of a set of computer codes ...

The safety assessment and licensing of nuclear reactor plants by the United States Nuclear Regulatory Commission (USNRC) depend partially on analytical computer programs to predict the response of safeguard systems to accident conditions. CONTEMPT4/MOD2 is a new computer code to predict the long-term thermal hydraulic behavior of water-cooled nuclear ...

The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA ...

MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ``MELCOR Verification, ...

MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor (LWR) nuclear power plants and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories. Brookhaven National Laboratory (BNL) has a program with the NRC called MELCOR Verification, ...

The SCDAP-3D computer code (Coryell 2001) has been developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of ...

The SCDAP-3D computer code (Coryell 2001) has been developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of ...

The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by ...

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ...

Coding theory applications to communications systems - code correlations, pseudorandom coded ranging

The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated ...

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system and the core during a severe accident transient, ...

Some years ago, within the framework of the study for the International Thermonuclear Experimental Reactor (ITER), ENEA assessed the RELAP5 code capability to simulate Helium cooled systems on the experimental data provided by the helium facility HEFUS3 (Brasimone, Italy). This activity allowed acquiring a certain experience on the limits and capabilities of the ...

The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) ...

RETRAN represents a new computer code approach for analyzing the thermal-hydraulic responses of Nuclear Steam Supply Systems to hypothetical Loss-of-Coolant-Accidents (LOCAs) and operational transients. In contrast to the conservative approach, RETRAN provides best estimate solutions to hypothetical LOCAs and operational transients. RETRAN is a computer ...

This report covers a segment of work done to develop methods for probabilistic quantification of uncertainties and identification of conservatisms in nuclear power plant safety analyses such as those performed for the loss-of-coolant accident. The process studied involves the use of best-estimate codes to determine the output distribution induced by input ...

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version ...

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, ...

This paper discusses the importance of modeling the transient behavior of multigroup cross sections in the context of coupled reactor physics and thermal-hydraulic computations with the SAS-DIF3DK computer code. The MACOEF macroscopic cross section methodology is presented. Results from benchmark verification calculations with a continuous-energy Monte Carlo are reported. ...

Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special ...

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...

... Title : Uncertainty and Sensitivity Analysis of a Fire-Induced Accident Scenario Involving Binary Variables and Mechanistic Codes. ...

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents ...

Safety analysis contractors responsible for existing nuclear facilities are required to submit a Documented Safety Analysis to the Department of Energy for approval by April 2003. Recognizing that schedule and resource limitations may be significant, an initiative is underway to make available a set of guidance tools. The guidance is in the form of a peer-reviewed Accident ...

Severe accident phenomena pertinent to the heavy-water-moderated production reactors of the US Department of Energy are being studied in the Severe Accident Analysis Program (SAAP) at the Savannah River Site. The SAAP has sought to define the behavior of the Savannah River reactors in accident scenarios involving significant fuel ...

The AP600 reactor core approaches buoyancy-dominated flow at the departure from nucleate boiling (DNB)-limiting period of a postulated steam-line--break accident. The reactor core has a highly skewed power distribution at this time due to the conservative assumption of a withdrawn rod cluster control assembly (stuck rod). Under such conditions, strong buoyancy-induced core ...

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in ...

of organizational accidents and to illustrate both effective and ineffective application of safety ..... Managing the Risks of Organizational Accidents, by ...

... Title : JET FIGHTER ACCIDENT/ATTRITION RATES IN PEACETIME: AN APPLICATION OF RELIABILITY GROWTH MODELLING,. ...

The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. ...

The models of the mechanistic code MFPR (Module for Fission Product Release) developed by IBRAE in collaboration with IRSN are described briefly in the first part of the paper. The influence of microscopic defects in the UO2 crystal structure on fission-gas transport out of grains and release from fuel pellets is described. These defects include point defects such as ...

Because of the complexity, volume of data and calculations required, one preferred analytical tool to perform transportation risk assessments is the RADTRAN computer code. RADTRAN combines user-determined material, packaging, transportation, demographic and meteorological factors, with health physics data to calculate expected radiological consequences and ...

According to the International Atomic Energy Agency (IAEA), industrial radiography accounts for approximately half of all reported accidents for the nuclear related industry. Detailed information about these accidents have been published by the IAEA in its Safety Report Series, one of which describes the radiological accident which ...

The TRAC-PD2 program provides capability for pressurized water reactors and best-estimate predictions of postulated accidents for many thermal-hydraulic test facilities. The code features a three-dimensional treatment of the pressure vessel and its associated internals; two-phase, nonequilibrium hydrodynamic models; flow-regime-dependent constitutive ...

A fuel motion detection system based on coded aperture imaging has been developed for the Annular Core Research Reactor. Its configuration evolved after investigations were carried out to determine the required system capabilities. The reactor environment, developments in the theory of coded apertures for nuclear radiations and compatibility with ...

An evaluation of ECCS performance for a loss-of-coolant accident in the Shippingport LWBR is presented. The application of the FLASH-6 code to LOCA analysis in the LWBR is also discussed.

The value of 0.34 gm/cm sup 3 for the minimum ink solute concentration is devised to protect the validity of the current applicable, generic Confinement Protection Limit (CPL) calculations. These CPL's are based on AA3 code accident calculations that use ...

AXAIRQ is a dose mode code used for prospective accident assessment at the Savannah River Site and is primarily used to show regulatory compliance. For completeness of pathway analysis, an ingestion model, AXINGST, has been developed for use with, and incorporation in, AXAIRQ. Currently available ingestion models were referenced as a basis for AXINGST. ...

As a part of the charter of the Severe Accident Sequence Analysis (SASA) Program, station blackout transients have been analyzed using a RELAP5 model of the Browns Ferry Unit 1 Plant. The task was conducted as a partial fulfillment of the needs of the US ...

The PRISM reactor is presently under pre-application licensing review by the NRC, with Brookhaven National Laboratory (BNL) providing technical assistance. The purpose of this paper is to review the current PRISM design and describe the results from the SSC Code calculations performed at BNL, for a series of unscrammed accidents. 3 ...

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal ...

The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNL...

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 publication, containing reports of U. S. Civil aircraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic...

The publication, contains reports of U. S. Civil aricraft accidents, in brief format, is issued monthly and includes those accidents analyzed and processed during the preceding month. The reports are reproduced directly from the coded record on magnetic t...