To carry out the accident analyses for the HANARO, RELAP5/MOD2 was modified. The heat transfer correlations and other thermal hydraulic models to account for the finned fuel elements and plate-type heat exchangers were modified to give RELAP5/KMRR, simula...

circulation studies The theoretical skills include � CFD code CFX � Thermal hydraulic system codes RELAP5-coolant interaction: COMETA � Thermal hydraulic system codes, severe accident and reactor core analyses tools: RELAP5 or national thermal hydraulic codes ...

An effort to parallelize RELAP5 was undertaken in order to capitalize on recent computational hardware developments in achieving the goal of realtime RELAP5 simulation. The objective was attained on a Stardent Titan 750 vector-parallel machine as proof of concept during a small break loss-of-coolant ...

RELAP5-3D is used worldwide for analyzing nuclear reactors under both operational transients and postulated accident conditions. Development of the RELAP code series began in 1975 and since that time the code has been continuously improved, enhanced, verified and validated [1]. Since RELAP5-3D ...

The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering ...

The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering ...

This report documents the analysis of a postulated Large Break Loss-of-Coolant Accident (LBOCA) in a Westinghouse 3-Loop PWR design using the 'Best Estimate' code RELAP5/MOD3. This LBLOCA calculation represents ENUSA's contribution to the 'Code Assessment...

A program was implemented as Savannah River site (SRS) to use the system code RELAP5 for K15.1 loss-of-coolant accident (LOCA) and loss-of-pumping accident (LOPA) reactor power limits calculations. The RELAP5 improvement program consolidated one system code and upgraded existing models. The existing SRS ...

The RELAP5 independent assessment project is part of an overall effort to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The RELAP5/MOD1 code has been as...

The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-n...

The analysis group of Almaraz Nuclear Power Plant has developed a model of the plant with RELAP 5/MOD.2/36:04. This model is the result of the work-experience on the code RELAP 5/MOD.1/ that was the standard code during the period 1984--1989. Different solutions were adopted in the network to implement the ...

Immediately after the accident in the fourth power unit of the Chernobyl nuclear power plant many computational and analytical investigations of the development of the accident and the values of separate factors in this situation were performed at the NIKIET and other research centers. Foreign research centers, where, in particular, the modern general-loop ...

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been ...

The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure tube ruptures. These ...

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 ...

In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core ...

SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to ...

The U.S. Nuclear Regulatory Commission has been conducting studies to evaluate the risk associated with steam generator tube failure during low probability severe accidents in pressurized-water reactors (PWRs) employing U-tube-type steam generators in par...

The RELAP5/MOD3 computer code has been developed for best-estimate simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents...

This paper presents experiment results from a simulated small-break loss-of-coolant accident (SBLOCA) involving natural circulation core heat rejection mechanisms. The experiment results are compared to pre-experiment RELAP5/MOD2 calculations, and finally...

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as p...

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 comparison to data showed that especially ...

Computational fluid dynamics (CFD) is used to predict steam generator inlet plenum mixing during a particular phase of a severe accident in a pressurized-water reactor. Boundary conditions are obtained from SCDAP/RELAP5 predictions of a TMLB station black...

In this paper a pressurized water reactor plant scenario is evaluated during a loss of feedwater following multiple failures and operator interventions. For a more realistic analysis, an experiment performed on the SPES facility is taken as a reference. A suitably qualified RELAP5/MOD2 code nodalization was used to evaluate the behavior of the facility ...

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A ...

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 ...

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-hydraulic system response to large break loss ...

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-hydraulic system response to large break loss ...

This report deals with a CORA-9 post-test-calculation with the computer-code RELAP5/SCDAP Mod.2.5 Ver.3f. It is the second technical report concerning the BMFT founded project ''Comparative Assessment of Different Computer Codes for Severe Accident Analys...

The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The ...

The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) ...

The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of ...

The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of ...

A new modified version of the RELAP5/MOD2 computer code for the analysis of reflood phase after a hypothetical large break loss-of-coolant accident was developed. Various rewetting correlations were examined and compared with FLECHT-SEASET reflood experimental data. The RELAP5 prediction of vapor temperatures was low in comparison with the data. The use of a new interfacial ...

The present study consists of the simulation of two loss of coolant accidents, LOCA 6' and LOCA 2', in one of the residual heat removal system (RHR) lines outside the containment, using the thermal-hydraulic code RELAP/MOD3.2. Both transients have been si...

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 product released during ...

The Small Break Loss of Coolant Accident (SBLOCA) Test Facility GERDA (Geradrohr Dampferzeuger Anlage) was designed to evaluate the post SBLOCA thermal-hydraulic events expected to occur in the German Mulheim-Karlich plant. This report presents the result...

The RELAP5 assessment project assessment project is part of an overall effort to determine the ability of various systems codes to predict the detailed thermal-hydraulic response of LWRs during accident and off-normal conditions. The RELAP5 code is being ...

RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling Sys...

This report presents an assessment study for the use of the code RELAP 5/MOD3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP 5/MOD1, was found adequate for this kind of calculations by EPRI. The hydrodynamic loads are very important for the ...

Current SCDAP/RELAP5 oxidation models have proven to under-predict oxidation, and therefore hydrogen production, when modeling reflood during in-pile tests. As an example, while OECD LOFT Experiment LP-FP-2 shows significant increases in temperature and pressure during reflood due to increased oxidation, only minimal additional oxidation is currently ...

Severe accident natural circulation flows have been investigated at the Idaho National Engineering Laboratory to better understand these flows and their potential impacts on the progression of a pressurized water reactor severe accident. Parameters affecting natural circulation in the reactor vessel and hot legs were identified and ranked based on their ...

As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety ...

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a ...

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ...

In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the ...

Benchmarking calculations utilizing RELAP5/MOD2.5 with a detailed multi-dimensional r-{theta} model of the SRS L-Reactor will be presented. This benchmarking effort has provided much insight into the two-component two-phase behavior of the reactor under isothermal conditions with large quantities of air ingested from the moderator tank ...

This article describes the development of an integrated accident analysis capability considering both reactor vessel and containment system responses. This integrated package, which uses the RELAP5 and CONTAIN computer codes, provides the user with greater accuracy and modeling flexibility when compared with accident analyses using these codes separately. ...

This study is part of the preliminary safety analysis for the new power expansion project on the University of Missouri Research Reactor (MURR). The loss of coolant accident (LOCA), which is initiated by hypothetical pipe ruptures at the most adverse positions (V507 A B) in both the hot and cold legs of the primary coolant loop, is analyzed with the thermohydraulic transient ...

Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles, and the steam generator (SG) ...

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) ...

RELAP5/MOD2 is a new version of RELAP5 containing improved modeling features that provide a generic pressurized water reactor transient simulation capability. In particular, the nonequilibrium modeling capability has been generalized to include post critical heat flux (CHF) conditions which occur in severe core damage accidents and ...

The advanced best-estimate systems codes currently being developed for the NRC are designed to provide realistic, rather than conservative, predictions of LWR plant behavior during a variety of accidents and transients. The RELAP5 independent code assessment project at Sandia National Laboratories is part of an overall code assessment effort funded by the NRC to arrive at a ...

An analysis of a loss-of-pumping accident (LOPA) in the Savannah River K-reactor has been performed using RELAP5. ECS was assumed to enter the system from only two systems in this case. No boiling occurred in the core during this calculation. Additional calculations were also performed in which the ECS delivery rate was reduced to ...

An effort to parallelize RELAP5 was undertaken in order to capitalize on recent computational hardware developments in achieving the goal of realtime RELAP5 simulation. The objective was attained on a Stardent Titan 750 vector-parallel machine as proof of...

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on ...

A large break loss of coolant accident in a CANDU can lead to degraded fuel cooling in a large number of fuel channels due to the apparition of a prolonged flow stagnation period in the downstream core pass. The paper presents a parametric study of a reactor inlet header break. The parametric survey includes: the size of the break, the choked flow model employed, the emergency ...

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating ...

A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B{sub 4}C, stainless steel, and Zircaloy. This ...

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 recognized experts to determine overall technical adequacy, even ...

The RELAP5/MOD2 (cycle 36.04) code is a one-dimensional, two-fluid, nonequilibrium, nonhomogeneous transient analysis code designed to simulate operational and accident scenarios in pressurized water reactors (PWRs). System models are solved using a semi-implicit finite difference method. The code was developed at EG and G in Idaho Falls under sponsorship of the US Nuclear ...

This paper reports on the RELAP5 computer code used to simulate four small-scale loss-of-coolant accident (LOCA) experiments. The purpose of the study is to help assess RELAP5 under conditions similar to those expected during a large-break LOCA at an Advanced Test Reactor (ATR). During an ATR large-break LOCA, it is expected that the ...

The main purpose of this report is to design a printed circuit heat exchanger (PCHE) for the Next Generation Nuclear Plant and carry out Loss of Coolant Accident (LOCA) simulation using RELAP5-3D. Helium was chosen as the coolant in the primary and secondary sides of the heat exchanger. The design of PCHE is critical for the LOCA ...

During the licensing process of a Canada deuterium uranium (CANDU) system, assessment of the system's thermohydraulic (T/H) code has been emphasized because the T/H analysis drives the other analyses, such as reactor physics, fuel, channel, and containment. Atomic Energy of Canada Limited has used the single-fluid code FIREBIRD for the system T/H analysis with conservative assumptions, ...

Inherent to modern-day analysis of thermal systems for nuclear power plants is the extensive use of thermal-hydraulic codes, such as TRAC, RETRAN, RELAP5/MOD2, etc., that quantitatively model individual components and their connectivity to represent a thermal-hydraulic system. The application of these codes is essential in the areas of design, safety, ...

The Simplified Boiling Water Reactor (SBWR) proposed by General Electric (GE) is an advanced light water reactor (ALWR) design that utilizes passive safety systems. The PCCS is a series of heat exchangers submerged in water and open to the containment. Since the containment is inerted with nitrogen during normal operation, the PCCS must condense the steam in the presence of ...

This report presents the results of the RELAP5/MOD2 assessment utilizing a Semiscale intermediate break of loss-of-coolant experiment S-IB-3. Comprehensive analysis with RELAP5/MOD2 is performed to predict the transient thermal-hydraulic responses of the experiment. Test S-IB-3 is a 21.7%, communicative cold leg break LOCA experiment using Semiscale Mod-2A facility in 1982, ...

This report presents the results of the RELAP5/MOD2 posttest assessment utilizing a Semiscale large break loss-of-coolant experiment numbered S-06-3. Test S-06-3 is a 200% double ended cold leg break experiment performed in Semiscale Mod-1 facility in 1978 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large LOCA in a ...

analysis. The most important are the versions of TRAC (Schnurr, et al., 1990), RELAP5 (1995), COBRA. The reflood model of COBRA-TF as well as other codes (RELAP5, TRAC, CATHARE) uses an axial fine mesh zone Fluid Mechanics and Heat Transfer, Taylor&Francis. #12;The ...

The University of Missouri Research Reactor (MURR) has been licensed to operate at 10 MW power since 1974. A preliminary study to increase its present power to a much higher level (/approx/30 MW) shows that this is readily achievable for steady-state operating conditions. The methods used to approach the goal of power upgrade operation include the flattening of the radial power distribution by ...

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor ...

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 ...

The USNRC version of the 3D neutron kinetics code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the USNRC thermal-hydraulic (T/H) codes RELAP5 and the consolidated TRAC (merger of TRAC-BF1 and TRAC-PF1). These coupled codes may be used to audit license safety analysis submittals where 3D spatial kinetics and thermal-hydraulic effects ...

In recent years, a new class of reactor designs has been proposed that utilize passive safety systems. General Electric has developed a Simplified Boiling Water Reactor (SBWR) design that relies on such passive systems. The SBWR has two passive cooling systems that involve energy transfer by condensation. These are the isolation condenser system (ICS) and the passive ...

A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and ...

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal ...

In this paper the SCDAP/RELAPS severe accident analysis computer code, developed at the Idaho National Engineering Laboratory, is used to analyze the fourth in a series of debris formation experiments. The debris formation-four (DF-4) experiment deals with heatup and meltdown of a boiling water reactor (BWR)-representative fuel and control blade assembly segment, performed in ...

Thermal-hydraulic system analysis in support of Savannah River Site reactor restart is being performed with a modified version of RELAP5/MOD2.5. This paper gives an overview of the Savannah River Site reactor system, the RELAP5 input models developed for the analysis, and the specific phenomena with which the code is having difficulty. The need for code development to address plenum phenomena, ...

Thermal-hydraulic transient/safety codes, such as RELAP, have been benchmarked for typical research reactor conditions, but such benchmarking for research reactors has not been adequate. Our simulation of the center fuel plate of the research reactor at the University of Missouri (MURR) using RELAP5/MOD2 after a hypothetical loss-of-coolant ...

The University of Missouri Research Reactor (MURR) has been licensed to operate at 10 MW power since 1974. A preliminary study to increase its present power to a much higher level (approx. 30 MW) shows that this is readily achievable for steady-state operating conditions. The methods used to approach the goal of power upgrade operation include the flattening of the radial power distribution by ...

This study was conducted to provide the basis for an estimate of the containment bypass risk presented by severe accident-induced steam generator tube rupture. During severe accidents, components can be heated to the extent that induced rupture might occur as a result of the combined effects of pressure and temperature. The potential for tube failure 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, core, fission product released during a severe accident transient as well as large and small break loss of coolant ...

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 loss of coolant ...

This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS ...

Reflooding of a hot damaged core following the start of a severe accident can lead to significant increases in the heating, melting, and oxidation of the core prior to the termination of the accident. These effects have been observed in bundle heating and melting experiments terminated by the addition of water and are postulated to have had a major impact ...

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate ...

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate ...

The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, ...

RELAP5/MOD3 is a reactor systems analysis code that has been developed jointly by the US Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that were members of the International Code A...

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, core, fission product released during a severe accident transient as well as large and small break loss of coolant ...

Analysis of OECD/CSNI ISP-42 Phase A PANDA Experiment using RELAP5/mod3.3 and GOTHIC 7.2a Code has been performed using RELAP5/mod3.3 and GOTHIC 7.2a code. For the RELAP5/mod33 calculation (ESBWR), GOTHIC, PANDA, Passive Containment Cooling System (PCCS), RELAP5, Wet Well ...

A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed ...

using the coupled system of codes RELAP5/PARCS, with models developed at the Department of Nuclear). The calculations were performed via the same RELAP5/PARCS model as the one presented in Section 3 for the Ringhals, J., St�lek, M. and Demazi�re, C. (2008) `Development of a coupled ...

In this work, the values calculated from two versions of the RELAP5/MOD1 code are compared with those measured in different tests. The first version of RELAP5 is the cycle 19 of the original version of INEL (RELAP5/MOD1-INEL) and the second version improv...

From 1975 to present, it has been found that the primary risk to the public health and safety from nuclear power reactors lies in ``beyond design basis'' accidents. During such severe accidents, melting of the reactor core may lead to a loss of primary system integrity, or even containment failure, which will allow escape of significant amounts of ...

A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with ...

An integrated code system consisting of RELAP5-3D and a multiphase CFD program has been created through the use of a generic semi-implicit coupling algorithm. Unlike previous CFD coupling work, this coupling scheme is numerically stable provided the material Courant limit is not violated in RELAP5-3D or at the coupling locations. The basis for the coupling scheme and details regarding the unique ...

An integrated code system consisting of RELAP5-3D and a multiphase CFD program has been created through the use of a generic semi-implicit coupling algorithm. Unlike previous CFD coupling work, this coupling scheme is numerically stable provided the material Courant limit is not violated in RELAP5-3D or at the coupling locations. The basis for the coupling scheme and details regarding the unique ...

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the ...

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the ...

The RELAP5YA computer code was developed to analyze postulated accidents and transients in light water reactor systems. The code has been assessed against many separate-effects and integral test results that address relevant thermal-hydraulic phenomena. The assessment results have established the validity of the code in predicting small- and large-break loss-of-coolant ...

Loss-of-coolant-accident (LOCA) and anticipated transient without scram (ATWS) calculations have been performed for the two Kernforschungszentrum Karlsruhe advanced pressurized water reactor reference designs (a homogeneous reactor with p/d = 1.2 and a heterogeneous reactor), for a homogeneous reactor with a tighter fuel rod lattice (p/d = 1.123), and for a reference ...

A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion ...

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the ...

The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. Each fuel assembly consists of a fuel rod bundle surrounded by a Zircaloy channel box. Each control blade consists of small stainless steel absorber tubes filled with B{sub 4}C powder surrounded by a stainless steel blade sheath. Under severe ...

The experiments QUENCH-01/06 were modelled using RELAP5/SCDAPSIM MOD3.2(bd) computer code. The results obtained from these models were compared to the experimental data to evaluate the code performance. The experiments were performed in the Forschungszentrum Karlsruhe (FZK), Germany. The objective of the experimental program was the investigation of the ...

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B&W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). ...

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). ...

This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design ...

The DPRA-SGTR computer program was written to develop a dynamic event tree for the analysis of a steam generator (SG) tube rupture (SGTR) event. Using the dynamic event tree, a full-scope understanding of the possible responses of a plant following an SGTR event and the related actions with the emergency operating procedures (EOPs) can be analyzed. ...

In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during LOCA, it will take many more resources to develop and to get final approval ...

An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce ...

The objective of this lower head creep rupture analysis is to assess the current version of MELCOR 1.8.5-RG against SCDAP/RELAP5 MOD 3.3kz. The purpose of this assessment is to investigate the current MELCOR in-vessel core damage progression phenomena including the model for the formation of a molten pool. The model for stratified molten pool natural heat ...

The thermal-hydraulic system computer code RELAP5/MOD2.5 was used to investigate the response of the primary cooling system during loss-of-coolant accidents (LOCAs) at the Savannah River Site (SRS) K-Reactor. In contrast to the conservative safety analyses performed to support the restart of K-Reactor, the assumptions and boundary ...

The thermal-hydraulic system computer code RELAP5/MOD2.5 was used to investigate the response of the primary cooling system during loss-of-coolant accidents (LOCAs) at the Savannah River Site (SRS) K-Reactor. In contrast to the conservative safety analyses performed to support the restart of K-Reactor, the assumptions and boundary ...

Argonne National Laboratory (ANL) performed audit calculations of a Reactor Coolant Pump (RCP) seized/sheared shaft transient for the Westinghouse Seabrook Plant using RELAP5/MOD 1.5 (Cycle 32) and FRAP-T6. The objective was to determine the effect of time of loss of offsite power and other single component failures on the peak clad temperature. The RCP ...

Research is being conducted to better understand natural circulation phenomena in mixtures of steam and noncondensibles and its influence on the temperature of the vessel internals and the hot leg, pressurizer surge line, and steam generator tubes. The temperature of these structures is important because their failure prior to reactor vessel lower head failure could reduce the likelihood of ...

Research is being conducted to better understand natural circulation phenomena in mixtures of steam and noncondensibles and its influence on the temperature of the vessel internals and the hot leg, pressurizer surge line, and steam generator tubes. The temperature of these structures is important because their failure prior to reactor vessel lower head failure could reduce the likelihood of ...

This paper presents the application of RELAP5 to the calculation of a Large Break (200% doubled-ended rupture) Loss-of-Colant-Accident (LBLOCA) at the reactor vessel inlet for the proposed Westinghouse AP600 design. A parametric calculation was also performed to determine effects of loss of a complete Emergency Core Cooling system (ECCS) train. These calculations were ...

This paper presents the application of RELAP5 to the calculation of a Large Break (200% doubled-ended rupture) Loss-of-Colant-Accident (LBLOCA) at the reactor vessel inlet for the proposed Westinghouse AP600 design. A parametric calculation was also performed to determine effects of loss of a complete Emergency Core Cooling system (ECCS) train. These calculations were ...

During the past years, a number of reduced-scale test facilities have been constructed to investigate the physical phenomena of transients or accidents occurring in nuclear power plants. Since the behavior of a nuclear power plant is complicated, it is quite impossible for a small-scaled facility to simulate all the physical phenomena during the transient process. But, by way ...

Unit 5 of the Zaporozh'ye Nuclear Power Plant (ZNPP5), equipped with a VVER-1000/320 4-loop reactor, has been modelled in detail using the RELAP5/MOD3.2 thermal-hydraulic system code. The 4-loop model affords a fidelity with ZNPP5 in terms of the system geometry such as the point of emergency core cooling (ECC) injection, for example. Both the reactor vessel and steam generators were ...

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of ...

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the ...

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system ...

Seven tight-lattice NEPTUN-III bottom-flooding experiments are analyzed by using the frozen version of RELAP5, RELAP5/MOD3.3/BETA. This work is part of the Paul Scherrer Institute (PSI) contribution to the High Performance Light Water Reactor (HPLWR) European Union project and aims at assessing the capabilities of the code to model the reflooding phenomena ...

In order to evaluate the safety of the RBMK-1000 reactor, one must study the behavior of the reactor in emergency situations at low power. In this case the core is filled with water, close to the saturation state, and disturbances producing steam in the core may for a positive void coefficient lead to an appreciable power increase. In this article, the authors consider the maximum projected ...

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at Oak Ridge National Laboratory. This paper deals with thermal-hydraulic analysis of ANSR`s cooling systems during nominal and transient conditions, with the major effort focusing upon the construction and testing of computer models of the reactor`s primary, secondary and reflector vessel cooling systems. The code ...

A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B&W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission`s (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure ...

A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure ...

One of the most important uses of dose assessment models in response to accidents at nuclear facilities is to help provide guidance to emergency response managers for identifying, and mitigating, the consequences of an accident once the accident has been ...

This volume synopsizes the investigation of contributing factors of accidents at crossings with flashing light and crossbuck warning devices. The data sources utilized, accident site investigation and methodology for the accident analysis are briefly disc...

with that letter, the Board believes that proper unmitigated or bounding accident analyses utilizing appropriately 19�22, 2002, visit to the site. Although the hazard and accident analyses for the CST DSA. Unmitigated Accident Analyses. The staff found that unmitigated accident analyses were ...

...AVIATION United States Aircraft Accidents Abroad § 102.11 ...expenses attendant upon an accident to United States aircraft can be paid. In...utilized in the case of an aircraft accident since the law which...

, Heating 7.2, Wims4, Dragon, Cobra, Eranos2.1, Relap 5, TRACE, STAR-CD, Fluent. #12;AWARDS Purdue University

Numerical experiments performed on a single instruction multiple data-pipeline vector parallel (SIMD-PVP) architecture computing machine, e.g., a CRAY X-MP/48, demonstrate that current nuclear reactor systems codes can be restructured for concurrent multiprocessing and show wall clock performance improvements of 1.5 to 3.0 on a 4-CPU machine, depending on plant model, problem type, and problem ...

The incidence of single vehicle fatal rollover accidents involving utility vehicles was examined from the FARS file using data from years 1978-1980. Vehicle type was separated into 14 categories and the number of fatal accidents and occupant fatalities wa...

Role of new safety systems applied in advanced WWER-1000 (passive residual heat removal system, SPOT and passive core flooding system, HA-2) in severe accident prevention is considered in the paper. The following typical beyond-design accidents (BDBAs) that essentially determine the design basis of the above passive systems are considered in the paper: - ...

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the ...

This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B&W) design (Oconee) and a Westinghouse (W) four-loop design ...

CERBERUS, a six-equation parallel thermal-hydraulic system simulation code, is being developed at the Idaho National Engineering Laboratory (INEL). CERBERUS Ver.00 performs parallel computations only for the heat transfer model. It is projected that CERBERUS Ver.01 will have a parallel heat transfer and hydraulic module, excluding the matrix solver, and CERBERUS Ver.02 will contain Ver.01 plus the ...

RELAP5 computer code was used to simulate an experiment designated 9.1.b, (2' Cold Leg Break without HPSI and with Delayed Ultimate Procedure) performed on BETHSY integral test facility. This test is characterized as beyond design transients scenarios wit...

RELAP5/MOD2 is a new version of RELAP5 containing improved modeling features that provide a generic pressurized water reactor transient simulation capability. In particular, the nonequilibrium modeling capability has been generalized to include post criti...

The Darlington NGS consists of four 940 MWe CANDU reactors. A detailed RELAP5 model of a Darlington NGS reactor has been created for use with RELAP5/MOD3.3. The model includes the primary heat transport system, the steam generator feedwater and main steam...

Assessment on the direct-contact condensation model was carried out using RELAP5/MOD2 Cycle 36.04 and the RELAP5/MOD3 Version 5m5 codes. The test data was obtained from the experiments at Northwestern University, which involved the horizontal cocurrent st...

, such as RELAP5 [1], used to predict axial void fraction variation during subcooled boiling are provided-fluid modeling codes (e.g., RELAP5) can be easily made. When using the correlations developed in this study received support from the United States Nuclear Regulatory Commission. References [1] RELAP5/MOD3 Code

Modifications and improvements to a previous RELAP5 model of the Vacuum Vessel Primary Heat Transfer System are described in this report. The modifications were new pump models, a new steam pressurizer, new coolant water control systems, additional pipe structures, and reduction of the pulse power to 6 MW.

of this capability, we have extracted the accumulator model from RELAP5 [6] into a separate application. Its direct/mpi2-report.html, (1997). [6] C. Allison (Ed.) et al., RELAP5/MOD3 Code manual, US Nuclear Regulatory

Calculations with the RELAP 5 computer code have been carried out for two LOCA-/BDHT-tests, run numbers 4903/16 and 6007/26, performed in the Two-Loop Test Apparaturs (TLTA) by GE. In this report RELAP 5-results are presented and compared with correspondi...

RELAP5/MOD2 simulations of post-dryout heat transfer in a 7 m long, 1.5 cm diameter heated tube are reported. The Biasi critical heat flux correlation is shown to be inadequate for predicting the experimental dryout. RELAP5 accurately predicted the measur...

. Existing thermal-hydraulic codes, used currently in industry [for instance RELAP5 (17)], have significant, Quebec, Canada, July 14-18, 2002. (17) RELAP5/MOD3 code manual, NUREG/CR-5535, 1995. (18) EUROPLEXUS code

reactor transient analysis code RELAP5/SCDAP was used in calculation. Rather detailed model of the plant investigation on RPV bypass flows and their influence on analyzed plant scenario is advised. Keywords-- RELAP5 will be given. II. OVERVIEW OF RELAP5/SCADAPSIM/MOD3.4 CODE The light water reactor transient ...

conditions for reactor safety analysis. The most important are the versions of TRAC [1], RELAP5 [2], COBRA processes are explicitly simulated in the code. The reflood model of COBRA-TF as well as other codes (RELAP5, Engineering Safety Analysis Group N-6, Los Alamos National Laboratory (November 1990). 2. The ...

...false How must I notify BLM of accidents occurring at my utilization facility...15 How must I notify BLM of accidents occurring at my utilization facility...You must orally inform us of all accidents that affect operations or...

This report summarizes the accidents and unscheduled events which may occur during the extraction, production, transportation, and utilization of non-nuclear energy technologies....

This report covers the social and economic effects of the accident at Three Mile Island during the first 6 months following the accident. A variety of data sources were utilized including published documents and statistics, household surveys, newspaper fi...

Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope w...

This report summarizes the accidents and unscheduled events which may occur during the extraction, production, transportation, and utilization of non-nuclear energy technologies.

...costs and benefits of severe accident mitigation design alternatives...for not incorporating severe accident mitigation design alternatives...utilizing the reactor in a nuclear power plant, or an evaluation...amendment either renders a severe accident mitigation design...

This note provides new information regarding the market reaction toward electric utility stocks that resulted both from the accident at Three Mile Island, and the events predating and postdating the accident. The results suggest that some of the market reaction heretofore ascribed to the accident resulted instead ...

A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed ...

A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed ...

The simplified boiling water reactor (SBWR) is a proposed 660-MW(electric) light water reactor plant designed by General Electric Company (GE). The SBWR is the newest of GE's advanced boiling water reactor (BWR) designs, incorporating natural circulation in the reactor coolant system and a host of passive safety features that allow for near-hands-off reactor safety during worst-case ...

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant ...

This paper presents experiment results from a simulated small-break loss-of-coolant accident (SBLOCA) involving natural circulation core heat rejection mechanisms. The experiment results are compared to pre-experiment RELAP5/MOD2 calculations, and finally are related to expected pressurized water reactor (PWR) behavior during SBLOCA ...

This book is the story of the TMI accident told by those who were there; the scientists and engineers at the utility operating the reactor and at the national laboratories that analyzed the accident. The first section describes the accident; the second focuses on the chemical aspects involved; and the third ...

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. The research will improve two well-respected light water reactor transient response codes ...

Significant multidimensional pressure gradients occur in the water plenum region of the reactors at the Savannah River Site (SRS). A multidimensional RELAP5 input model of the L-Reactor was developed and benchmarked against SRS data. Although RELAP5 is a one-dimensional code, its cross-flow junction allows a ...

A new model has been built into RELAP5/MOD3 to facilitate coupling RELAP5/MOD3 and other computer codes. The new model has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 {open_quotes}styled{close_quotes} input and the Parallel Virtual Machine ...

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain ...

The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating void fraction distribution over height in RBMK fuel channels performed in Facility BM at the ENTEK. This is RBMK Standard Problem 7 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian Minatom ...

Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are ...

Jun 16, 2011 ... Integrated management of carbon sequestration and biomass utilization ... in the Korean War and I subsequently became a political accident. ...

One of the major improvements in the progress from RELAP5/MOD2 to RELAP5/MOD3 was the modification to improve portability. The use of the code was thus extended from the 60 and 64 bit work mainframe computers to 32 bit workstations and PC's. This has take...

the energy balance on the interface between the phases (Relap5, 1995) )()( ,, ssatlssatg ilig ig phph QQ coefficients are calculated by applying the models of Relap5 for bubbly flow into water pool and into large

removal system using RELAP5-3D. � Fusion neutron source calculated from plasma particle and power balance, coupled to the point neutron kinetics equation through source term. � ATHENA version of RELAP5-3D allows

Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonpo...

Evaluation of the accuracy of large thermohydraulic codes and the safety margins of light water reactors are among the objectives of international research programs such as those organized by the Committee on the Safety of Nuclear Installations (CSNI) and...

A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geometric...

The RELAP5(star) Electronic Newsletter Service is a subscriber-funded information service provided by EG&G Idaho, Inc., to authorized recipients of the RELAP5/MOD2 computer code. This computer-based, electronic Newsletter Service contains technical inform...

The RELAP5 independent code assessment project at Sandia National Laboratories is part of an overall code assessment effort funded by the NRC to arrive at a qualified judgment of the accuracy with which these codes can predict the detailed thermal/hydraul...

This report describes the implementation of the PVM API in the RELAP5-3D© computer code. The information in the report is intended for programmers wanting to correct or extend RELAP5-3D©.

This report describes the implementation of the PVM API in the RELAP5-3D� computer code. The information in the report is intended for programmers wanting to correct or extend RELAP5-3D�.

RELAP5/MOD3 is a pressurized water reactor (PWR) system analysis code being developed jointly by the US Nuclear Regulatory Commission (USNRC) and consisting of several of the countries that are members of the International Code Assessment and Applications...

The effects of reactor coolant system natural circulation on the response of the Surry Nuclear Power Plant during a station blackout transient were investigated. A TMLB' sequence (loss of all ac power, immediate loss of auxiliary feedwater) was simulated ...

The RELAP5-3D computer code was modified to make the explicit coupling capability in the code fully functional. As a test of the modified code, a coupled RELAP5/RELAP5 analysis of the Edwards-O'Brien blowdown problem was performed which showed no significant deviations from the standard ...

One of the major improvements in the progress from RELAP5/MOD2 to RELAP5/MOD3 was the modification to improve portability. The use of the code was thus extended from the 60 and 64 bit work mainframe computers to 32 bit workstations and PC`s. This has taken RELAP5 from a few types of mainframes` environmentally ...

High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the world�?s energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential ...

attuale dei codici termoidraulici di sistema (come relap5, Cathare2, Trace,..), � basata sulla soluzione in the safety evaluation of nuclear reactors. The actual generation of system thermal-hydraulic codes (e.g Relap like: Relap 5, Cathare 2, Trace. The Cathare2 code developed at Commissariat a l'Energie ...

52 B1.1. RELAP-5 52 B1.2. TRAC(PP1) 54 B1.3. NORA 60 B2. Comparison with data 61 B3. Experiences surface using heat transfer package of RELAP-5 with default CHF- correlation 55 #12;- 5 - Bl.2. As B1. length (NORA-case 1) 81 B2.22. Temperature vs. length (RELAP-5 case 1) 82 B2.23. Temperature vs. length

The present work reports comparisons between experimental and theoretical data done with the RELAP5/MOD1 and TRAC-PD2 codes, with particular emphasis on RELAP5/MOD1 code run with basic experimental data from the CANON depressurization simulation. This exp...

As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of this computer program versus a natural ...

The NRC version of the 3-D neutron kinetics code PARCS was coupled with the NRC thermal-hydraulics codes RELAP5 and TRAC-M. The MSLB benchmark problem was performed to provide a consistent assessment of the paired codes. Results were presented using the return to power cross section set provided in the benchmark problem. The two code pairings gave similar ...

A multidisciplinary approach to accident investigation is routine in commercial aircraft accidents. A medical team is usually assigned to assist with the investigation. As far as general aviation accidents are concerned, in a majority of cases, these resources are not available to the investigator-in-charge. We describe a general ...

Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. ...

The reactor design features 4 independent cooling loops (3 active, 1 standby), each containing a main circulation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and also that the check valve in that loop remains stuck open. This accident is considered extremely unlikely. Flow ...

A second-order accurate scheme based on high-resolution shock-capturing methods was used with a typical two-phase flow model which is used in the computer codes for simulation of nuclear power plant accidents. The two-fluid model, which has been taken from the computer code RELAP5, consists of six first-order partial differential ...

A brief historical background and a description of short- and long-term task plan development for effective closure of this important safety issue for the HFIR are given. Short-term aspects deal with Fuel-Coolant-Interaction (FCI) issues experimentation, modeling, and analysis for the flow-blockage-induced steam explosion events in direct support of the SAR. Long-term aspects deal with addressing ...

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a LOCA, it generally takes much more resources to develop. ...

The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). ...