Development and application of the PCRELAP5 - Data Calculation Program for RELAP 5 Code

Energy Technology Data Exchange (ETDEWEB)

Silvestre, Larissa J.B.; Sabundjian, Gaianê, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Cálculo do RELAP5 – PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. An English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. The final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra-2. (author)

Data calculation program for RELAP 5 code

International Nuclear Information System (INIS)

Silvestre, Larissa J.B.; Sabundjian, Gaiane

2015-01-01

As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)

Data calculation program for RELAP 5 code

Energy Technology Data Exchange (ETDEWEB)

Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)

Relap4/SAS/Mod5 - A version of Relap4/Mod 5 adapted to IPEN/CNEN - SP computer center

International Nuclear Information System (INIS)

Sabundjian, G.

1988-04-01

In order to improve the safety of nuclear reactor power plants several computer codes have been developed in the area of thermal - hydraulics accident analysis. Among the public-available codes, RELAP4, developed by Aerojet Nuclear Company, has been the most popular one. RELAP4 has produced satisfactory results when compared to most of the available experimental data. The purposes of the present work are: optimization of RELAP4 output and messages by writing there information in temporary records, - display of RELAP4 results in graphical form through the printer. The sample problem consists on a simplified model of a 150 MW (e) PWR whose primary circuit is simulated by 6 volumes, 8 junctions and 1 heat slab. This new version of RELAP4 (named RELAP4/SAS/MOD5) have produced results which show that the above mentioned purposes have been reached. Obviously the graphical output by RELAP4/SAS/MOD5 favors the interpretation of results by the user. (author) [pt

SCDAP/RELAP5/MOD3 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

1991-01-01

The SCDAP/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 Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

International Nuclear Information System (INIS)

Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

2013-01-01

The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

Pump-stopping water hammer simulation based on RELAP5

International Nuclear Information System (INIS)

Yi, W S; Jiang, J; Li, D D; Lan, G; Zhao, Z

2013-01-01

RELAP5 was originally designed to analyze complex thermal-hydraulic interactions that occur during either postulated large or small loss-of-coolant accidents in PWRs. However, as development continued, the code was expanded to include many of the transient scenarios that might occur in thermal-hydraulic systems. The fast deceleration of the liquid results in high pressure surges, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increase. This phenomenon is called water hammer. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the system when the pressure surges become considerably high. If this happens and when the pressure exceeds the critical pressure that the pipe or the fittings along the pipeline can burden, it will result in the failure of the whole pipeline integrity. The purpose of this article is to introduce the RELAP5 to the simulation and analysis of water hammer situations. Based on the knowledge of the RELAP5 code manuals and some relative documents, the authors utilize RELAP5 to set up an example of water-supply system via an impeller pump to simulate the phenomena of the pump-stopping water hammer. By the simulation of the sample case and the subsequent analysis of the results that the code has provided, we can have a better understand of the knowledge of water hammer as well as the quality of the RELAP5 code when it's used in the water-hammer fields. In the meantime, By comparing the results of the RELAP5 based model with that of other fluid-transient analysis software say, PIPENET. The authors make some conclusions about the peculiarity of RELAP5 when transplanted into water-hammer research and offer several modelling tips when use the code to simulate a water-hammer related case

Pump-stopping water hammer simulation based on RELAP5

Science.gov (United States)

Yi, W. S.; Jiang, J.; Li, D. D.; Lan, G.; Zhao, Z.

2013-12-01

RELAP5 was originally designed to analyze complex thermal-hydraulic interactions that occur during either postulated large or small loss-of-coolant accidents in PWRs. However, as development continued, the code was expanded to include many of the transient scenarios that might occur in thermal-hydraulic systems. The fast deceleration of the liquid results in high pressure surges, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increase. This phenomenon is called water hammer. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the system when the pressure surges become considerably high. If this happens and when the pressure exceeds the critical pressure that the pipe or the fittings along the pipeline can burden, it will result in the failure of the whole pipeline integrity. The purpose of this article is to introduce the RELAP5 to the simulation and analysis of water hammer situations. Based on the knowledge of the RELAP5 code manuals and some relative documents, the authors utilize RELAP5 to set up an example of water-supply system via an impeller pump to simulate the phenomena of the pump-stopping water hammer. By the simulation of the sample case and the subsequent analysis of the results that the code has provided, we can have a better understand of the knowledge of water hammer as well as the quality of the RELAP5 code when it's used in the water-hammer fields. In the meantime, By comparing the results of the RELAP5 based model with that of other fluid-transient analysis software say, PIPENET. The authors make some conclusions about the peculiarity of RELAP5 when transplanted into water-hammer research and offer several modelling tips when use the code to simulate a water-hammer related case.

Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

SCDAP/RELAP5 independent peer review

Energy Technology Data Exchange (ETDEWEB)

Corradini, M.L. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering; Dhir, V.K. [Dhir, (V.K.) Santa Monica, CA (United States); Haste, T.J. [AEA Technology, Winfrith (United Kingdom); Heames, T.J. [Science Applications, Inc., Albuquerque, NM (United States); Jenks, R.P. [Los Alamos National Lab., NM (United States); Kelly, J.E. [Sandia National Labs., Albuquerque, NM (United States); Khatib-Rahbar, M. [Energy Research, Inc., Rockville, MD (United States); Viskanta, R. [Purdue Univ., Lafayette, IN (United States). Heat Transfer Lab.

1993-01-01

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 though the code is still under development. For this purpose, an eight-member SCDAP/RELAP5 Peer Review Committee was organized, and the outcome of the review should help the NRC prioritize future code-development activity. Because the code is designed to be mechanistic, the Committee used a higher standard for technical adequacy than was employed in the peer review of the parametric MELCOR code. The Committee completed its review of the SCDAP/RELAP5 code, and the findings are documented in this report. Based on these findings, recommendations in five areas are provided: (1) phenomenological models, (2) code-design objectives, (3) code-targeted applications, (4) other findings, and (5) additional recommendations.

SCDAP/RELAP5 independent peer review

International Nuclear Information System (INIS)

Corradini, M.L.; Haste, T.J.; Heames, T.J.; Jenks, R.P.; Kelly, J.E.; Khatib-Rahbar, M.; Viskanta, R.

1993-01-01

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 though the code is still under development. For this purpose, an eight-member SCDAP/RELAP5 Peer Review Committee was organized, and the outcome of the review should help the NRC prioritize future code-development activity. Because the code is designed to be mechanistic, the Committee used a higher standard for technical adequacy than was employed in the peer review of the parametric MELCOR code. The Committee completed its review of the SCDAP/RELAP5 code, and the findings are documented in this report. Based on these findings, recommendations in five areas are provided: (1) phenomenological models, (2) code-design objectives, (3) code-targeted applications, (4) other findings, and (5) additional recommendations

RELAP5 - a new tool for pressurized water reactor safety analysis

International Nuclear Information System (INIS)

Perneczky, L.

1988-11-01

The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

Waste Evaporator Accident Simulation Using RELAP5 Computer Code

International Nuclear Information System (INIS)

POLIZZI, L.M.

2004-01-01

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 corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

International Nuclear Information System (INIS)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

2005-01-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

Energy Technology Data Exchange (ETDEWEB)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

2005-07-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

International Nuclear Information System (INIS)

Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane

2015-01-01

Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

Energy Technology Data Exchange (ETDEWEB)

Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

RELAP5/MOD3 code manual. Volume 4, Models and correlations

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best-estimate transient 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 and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations

Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

Energy Technology Data Exchange (ETDEWEB)

Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

1995-09-01

This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

Nonequilibrium constitutive models for RELAP5/MOD2

International Nuclear Information System (INIS)

Lin, J.C.; Trapp, J.A.; Riemke, R.A.; Ransom, V.H.

1983-01-01

RELAP5/MOD2 is a new version of RELAP5 containing improved modeling features that provide a generic pressurized-water transient simulation capability. In particular, the nonequilibrium modeling capability has been generalized to include conditions that occur in operational transients including repressurization and emergency-feed-water injection with loss-of-coolant accidents. The improvements include addition of a second energy equation to the hydrodynamic model, addition of nonequilibrium heat-transfer models, and the associated nonequilibrium vapor-generation models. The objective of this paper is to describe these models and to report the developmental assessment results obtained from similar of several separate effects experiments. The assessment shows that RELAP5 calculated results are in good agreement with data and the nonequilibrium phenomena are properly modeled

Simulation of the postulated stopping accident of the bombs of the primary circuit of Angra 2 with the code RELAP5/MOD3.2

International Nuclear Information System (INIS)

Andrade, Delvonei A.; Sabundjian, Gaiane; Madeira, Alzira A.; Pereira, Luiz Carlos M.; Borges, Ronaldo C.; Lapa, Nelbia S.

2001-01-01

This work presents the simulation of an anticipated transient for Angra 2 Nuclear Power Plant, where the coast down of the four reactor coolant pumps is verified. The best estimate thermal hydraulic system code RELAP5/MOD3.2 was used on this frame. A multi-purpose nodalization of Angra 2 was developed to simulate a comprehensive set of operational transients and accidents with RELAP5/MOD3.2 code. The overall objective of this work is to provide independent accident evaluation and further operational behavior follow-up to support the licensing process of the plant. (author)

SCDAP/RELAP5 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.

1996-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

Recent SCDAP/RELAP5 code applications and improvements

International Nuclear Information System (INIS)

Harvego, E.A.; Ghan, L.S.; Knudson, D.L.; Siefken, L.J.

1998-01-01

This paper summarizes (1) a recent application of the severe accident analysis code SCDAP/RELAP5/MOD3.1, and (2) development and assessment activities associated with the release of SACDAP/RELAP5/MOD3.2. The Nuclear Regulatory Commission (NRC) has been evaluating the integrity of steam generator tubes during severe accidents. MOD3.1 has been used to support that evaluation. Studies indicate that the pressurizer surge line will fail before any steam generator tubes are damaged. Thus, core decay energy would be released as steam through the surge line and the tube wall would be spared from exposure to prolonged flow of high temperature steam. The latest code version, MOD3.2, contains several improvements to models that address both the early phase and late phase of a severe accident. The impact of these improvements to the overall code capabilities has been assessed. Results of the assessment are summarized in this paper

Analysis of the SL-1 Accident Using RELAPS5-3D

International Nuclear Information System (INIS)

Francisco, A.D.; Tomlinson, E.T.

2007-01-01

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 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 excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with a discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty)

Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

International Nuclear Information System (INIS)

Yang, C. Y.; Bang, Y. S.; Kim, H. J.

2001-01-01

A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

Simulation of the IAEA's fourth Standard Problem Exercise small-break loss-of-coolant accident using RELAP5/MOD.3.1

International Nuclear Information System (INIS)

Cebull, P.P.; Hassan, Y.A.

1995-01-01

A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydraulic code. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). Blind calculations of the exercise are presented, and the timing of various events throughout the transient is discussed. A posttest analysis is performed in which the sensitivity of the calculated results is investigated. The code RELAP5 predicts most of the transient events well, although a few problems are noted, particularly the failure of RELAP5 to predict dryout in the core even through the collapsed liquid level fell below the top of the heated portion. A discrepancy between the predicted primary mass inventory distribution and the experimental data is identified. Finally, the primary and secondary pressures calculated by RELAP5 fell too rapidly during the latter part of the transient, resulting in rather large errors in the predicted timing of some pressure-actuated events

RELAP5 assessment: conclusions and user guidelines

International Nuclear Information System (INIS)

Kmetyk, L.N.

1984-10-01

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-normal conditions. The RELAP5/MOD1 code has been assessed at Sandia against a variety of test data from both integral and separate effects test facilities. All these analyses have been documented in detail in individual topical reports; in this paper we attempt to evaluate the overall code performance by comparing results from many different calculations, and to offer other users some guidelines based on our experience to date

Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

International Nuclear Information System (INIS)

Griggs, D.P.

1993-03-01

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 six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

International Nuclear Information System (INIS)

Coryell, E.W.

1995-06-01

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 accidents, operational transients such as anticipated transient without SCRAM, loss of off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models which are resident in the RELAP5 portion of the code. A description of the organization and structure of SCDAP/RELAP5 is presented. Additional information is provided regarding the manner in which models in one portion of the code impact other parts of the code, and models which are dependent on and derive information from other subcodes

Uncertainty Analysis of RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Alexandra E Gertman; Dr. George L Mesina

2012-07-01

As world-wide energy consumption continues to increase, so does the demand for the use of alternative energy sources, such as Nuclear Energy. Nuclear Power Plants currently supply over 370 gigawatts of electricity, and more than 60 new nuclear reactors have been commissioned by 15 different countries. The primary concern for Nuclear Power Plant operation and lisencing has been safety. The safety of the operation of Nuclear Power Plants is no simple matter- it involves the training of operators, design of the reactor, as well as equipment and design upgrades throughout the lifetime of the reactor, etc. To safely design, operate, and understand nuclear power plants, industry and government alike have relied upon the use of best-estimate simulation codes, which allow for an accurate model of any given plant to be created with well-defined margins of safety. The most widely used of these best-estimate simulation codes in the Nuclear Power industry is RELAP5-3D. Our project focused on improving the modeling capabilities of RELAP5-3D by developing uncertainty estimates for its calculations. This work involved analyzing high, medium, and low ranked phenomena from an INL PIRT on a small break Loss-Of-Coolant Accident as wall as an analysis of a large break Loss-Of- Coolant Accident. Statistical analyses were performed using correlation coefficients. To perform the studies, computer programs were written that modify a template RELAP5 input deck to produce one deck for each combination of key input parameters. Python scripting enabled the running of the generated input files with RELAP5-3D on INL’s massively parallel cluster system. Data from the studies was collected and analyzed with SAS. A summary of the results of our studies are presented.

SCDAP/RELAP5/MOD3 code development

International Nuclear Information System (INIS)

Allison, C.M.; Siefken, J.L.; Coryell, E.W.

1992-01-01

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 US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

Simulation and verification studies of reactivity initiated accident by comparative approach of NK/TH coupling codes and RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Ud-Din Khan, Salah [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center; Peng, Minjun [Harbin Engineering Univ. (China). College of Nuclear Science and Technology; Yuntao, Song; Ud-Din Khan, Shahab [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; Haider, Sajjad [King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center

2017-02-15

The objective is to analyze the safety of small modular nuclear reactors of 220 MWe power. Reactivity initiated accidents (RIA) were investigated by neutron kinetic/thermal hydraulic (NK/TH) coupling approach and thermal hydraulic code i.e., RELAP5. The results obtained by these approaches were compared for validation and accuracy of simulation. In the NK/TH coupling technique, three codes (HELIOS, REMARK, THEATRe) were used. These codes calculate different parameters of the reactor core (fission power, reactivity, fuel temperature and inlet/outlet temperatures). The data exchanges between the codes were assessed by running the codes simultaneously. The results obtained from both (NK/TH coupling) and RELAP5 code analyses complement each other, hence confirming the accuracy of simulation.

Utilization of the RELAP4 in the experimental water loop of the IPEN-CNEN/SP

International Nuclear Information System (INIS)

Sabundjian, G.

1991-09-01

For a better security of the nuclear power station and the population, thermal hydraulics codes for accident analysis were developed. Within all codes, the RELAP4, developed by Aerojet Nuclear Company of the Idaho Falls Engineering Laboratory (U.S.A.), has been the more used mostly by the fact of been available to the public. These code presents the most satisfactory results when used for various sort of problems. Basely the paper analyzes and compares the RELAP4/MOD3 and MOD5 versions, through the simulation on the Water Experimental Loop (C.E.A.) of the IPEN, in the steady state and in accidents conditions. Although the C.E.A. had been constructed to simulate PWR and BWR reactors, in these paper the tests were made only for PWR. Through the results analysis found in these paper, we concluded that the model adaptation for the simulations made with the RELAP4 code (version MOD3 and MOD5), were satisfactory. The accidents simulations indicates that the C.E.A. can support accident situation with the emergency system acting. Comparing the two versions, the MOD5 presented the better results compared with the real estimations. (author)

Improving containment mass and energy releases for CONTEMPT-LT/028 TU with RELAP5/MOD3

International Nuclear Information System (INIS)

DaSilva, H.C.; Choe, W.G.

1996-01-01

In order to obtain boundary conditions for RELAP5/MOD3 best estimate (BE) large break (LB) loss-of-coolant accident (LOCA) calculations, it is necessary to utilize a separate containment analysis code CONTEMPT-LT/028 TU, which in turn accepts mass and energy releases from the RELAP5/MOD3 calculation. When these boundary conditions are obtained, they are observed to be significantly lower than those reported in FSAR containment analyses. This motivates the present study, where RELAP5/MOD3 mass and energy releases are generated using the same assumptions listed in the FSAR containment calculations. Then CONTEMPT-LT/028 TU pressures and temperatures calculated with both sets of mass and energy releases are compared. It is seen that those obtained with the RELAP5/MOD3 input are still significantly lower, indicating a level of conservatism in the FSAR mass and energy releases that is even above that explicitly listed and also incorporated into the RELAP5/MOD3 calculation. An important conclusion from this finding is that Environmental Qualification (EQ) issues requiring containment re-analyses are likely to be easily resolved if new mass and energy releases are calculated with state-of-the-art LOCA codes modeling the entire reactor coolant system, even when conservative assumptions are incorporated

Comparison between RELAP5 versions for a two-phase natural circulation analysis

Energy Technology Data Exchange (ETDEWEB)

Braz Filho, Francisco A.; Ribeiro, Guilherme B.; Sabundjian, Gaianê; Caldeira, Alexandre D., E-mail: fbraz@ieav.cta.br, E-mail: gbribeiro@ieav.cta.br, E-mail: alexdc@ieav.cta.br, E-mail: gdjian@ipen.br [Instituto de Estudos Avançados (IEAv), São José dos Campos, SP (Brazil). Div. de Energia Nuclear; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

RELAP5 is one of the most used numerical tools to predict thermal-hydraulic and neutronic phenomena in nuclear reactors. RELAP5-3D is the latest version of this software family, but RELAP5-mod3 is still widely used in Brazilian research institutes and it is also used as benchmark for several nuclear applications. Among these applications, the use of passive heat transfer mechanisms, such as natural circulation, has drawn attention of several studies, especially after the Fukushima-Daiichi accident. Considering this aforementioned aspect, this study proposes a comparison of RELAP5-3D and RELAP5-mod3 versions, focusing on a two-phase natural circulation loop. For comparison purposes, an experimental data set is part of the analysis. Results showed that during the single-phase regime, the temperature difference between versions is negligible. However, when the two-phase flow regime takes place, different wavelengths and amplitudes of flow instabilities were obtained for each version. When compared to the experimental data set, the RELAP5-3D version provided the best prediction results. (author)

SCDAP/RELAP5 lower core plate model

International Nuclear Information System (INIS)

Coryell, E.W.; Griffin, F.P.

1999-01-01

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 core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled ''Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model''

SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Johnsen, E.C. [eds.; Allison, C.M. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1995-06-01

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 a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported.

SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

International Nuclear Information System (INIS)

Hohorst, J.K.; Johnsen, E.C.

1995-06-01

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 a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported

RELAP/MOD3 code manual: User's guidelines. Volume 5, Revision 1

International Nuclear Information System (INIS)

Fletcher, C.D.; Schultz, R.R.

1995-08-01

The RELAP5 code has been developed for best estimate transient 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, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume V contains guidelines that have solved over the past several years through the use of the RELAP5 code

Coupled fuel performance and thermal-hydraulics simulation with BISON and RELAP-7 at accident conditions

Energy Technology Data Exchange (ETDEWEB)

Martineau, R.C., E-mail: Richard.Martineau@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

2015-07-01

'Full text:' RELAP-7 is expected to be the next in the RELAP nuclear reactor safety/systems analysis application series developed at the Idaho National Laboratory (INL). The development of RELAP-7 began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support modern nuclear power safety analysis. RELAP-7 is built using the INL's modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved implicit numerical schemes, including higher-order integration in both space and time, and yielding converged second-order accuracy for RELAP-7. The code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. This component-based software architecture allows RELAP-7 to quickly adopt different physical models for different applications. One of the main advantages of building RELAP-7 on the MOOSE framework is that tight coupling with other MOOSE-based applications solving physics not present in RELAP-7 requires little to no additional lines of code. For example, the RELAP-7 core channel component is based upon a one-dimensional flow channel and a three-zone two-dimensional heat structure designed to represent fuel, gap, and cladding conjugate heat transfer with the coolant. However, the RELAP-7 application does not carry the fuels performance physics to analyze irradiated fuel, especially for accident scenarios. Here, we demonstrate the tightly coupled capability of the BISON nuclear fuels performance application with RELAP-7 for the station black out (SBO) accident scenario (Fukushima type event) and

RELAP5 model for TRIGA 14 MWt

International Nuclear Information System (INIS)

Negut, Gheorghe; Prisecaru, Ilie)

2003-01-01

The ICN TRIGA facility was commissioned at the beginning of 1980. Since that time the 14 MW Material Test Reactor was used extensively for various tests, experiments and basic research. There were provided a 100 kW loop and natural convection capsules to test CANDU type fuel and structural materials as Zircaloy as well as medical and industrial radioisotopes production facilities. The first load of High Enriched Uranium (HEU) fuel, mostly was exhausted and in the '90 there was necessary a replacement with Low Enriched Uranium (LEU) fuel. The original configuration of 29 HEU fuel bundles is now replaced with a HEU - LEU mixed core of 35 fuel bundles. This process involved the revision of Safety Analysis Report.The paper presents the analysis of Loss of Fluid Accident (LOFA) and the comparison with the results obtained during commissioning phase. A simple model of the TRIGA core was developed with the aid of RELAP5MOD3.2 code. The RELAP5 documented the flow reversal and natural convection establishment, and the model proved a useful and accurate instrument for thermal hydraulic analysis. Presented are the RELAP5 model for the TRIGA reactor and a LOFA accident analysis. The following results and conclusions concerning the LOFA tests with emergency pump off after 15 minutes and at start of the test are presented. In the first test TRIGA reactor is operated at the nominal power of 14 MW and the main pumps flow rate is 700 kg/sec. The main pumps are stopped and in 10 seconds the flow rate reaches 22 kg/sec, the emergency pump flow rate. The emergency pump is stopped after 15 minutes from the LOFA test initiation. The reactor is tripped by the low flow rate signal at the level of 473 kg/sec. The core flow is reversed after the 5 seconds and the core is cooled by natural convection. After 425 seconds from the LOFA initiation the residual power level is sufficiently low, so, the flow is reversed once again and the residual heat is removed by the emergency pump flow in

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3DC

International Nuclear Information System (INIS)

Knudson, D.L.; Rempe, J.L.

2002-01-01

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D C has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D C relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D C are outlined. (authors)

Angra 2 small break LOCA flow regime identification through RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Rocha, Marcelo da Silva; Sabundjian, Gaiane; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Torres, Walmir Maximo; Conti, Thadeu das Neves; Macedo, Luiz Alberto; Umbehaun, Pedro Ernesto; Mesquita, Roberto Navarro de; Masotti, Paulo Henrique Ferraz, E-mail: msrocha@ipen.br, E-mail: gdjian@ipen.br, E-mail: abelchior@ipen.br, E-mail: delvonei@ipen.br, E-mail: wmtorres@ipen.br, E-mail: tnconti@ipen.br, E-mail: lamacedo@ipen.br, E-mail: umbehaun@ipen.br, E-mail: s, E-mail: rnavarro@ipen.br, E-mail: pmasotti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2012-07-01

The purpose of this paper is to identify the flow regimes in the core of Angra 2 nuclear reactor with RELAP5/MOD3.2.gamma code (RELAP5, 2001). The postulated accident is the loss of coolant through a small break in the primary circuit (SBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 - FSAR (ETN, 2006). As the primary circuit pressure decreases due to the loss of coolant, several alternating two phase flow regimes are established in the primary circuit. This paper analyses the coolant two-phase flow behavior in the nuclear reactor core during the postulated accident. (author)

SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

International Nuclear Information System (INIS)

Hagrman, D.T.

1995-06-01

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 accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Directory of Open Access Journals (Sweden)

Andrej Prošek

2013-01-01

Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

RELAP/MOD3 code manual: User`s guidelines. Volume 5, Revision 1

Energy Technology Data Exchange (ETDEWEB)

Fletcher, C.D.; Schultz, R.R. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1995-08-01

The RELAP5 code has been developed for best estimate transient 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, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume V contains guidelines that have solved over the past several years through the use of the RELAP5 code.

RELAP5 simulation for one and two-phase natural circulation phenomenon

International Nuclear Information System (INIS)

Sabundjian, Gaiane; Andrade, Delvonei Alves de; Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Castro, Alfredo Jose Alvim de; Braz Filho, Francisco A.; Borges, Eduardo Madeira; Damy. Osvaldo Luiz Almeida; Torres, Eduardo

2007-01-01

The objective of this paper is to study the natural circulation phenomenon in one and two-phase regime. There has been a crescent interest in the scientific community in the study of the natural circulation. New generation of compact nuclear reactors uses the natural circulation for residual heat removal in case of accident or shutdown. For this study, the modeling and the simulation of the experimental circuit is performed with the RELAP5 code. The experimental circuit is mounted in the Chemical Engineering Department of the University of Sao Paulo. It is presented in this work the theoretical/experimental comparison for one and two-phase flow. These results will be stored in a database to validate RELAP5 calculations. This work was also used to training some users of RELAP5 from IEAv. (author)

Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

International Nuclear Information System (INIS)

Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

2004-01-01

The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

RELAP5/MOD2 calculation of OECD LOFT test LP-FW-01

International Nuclear Information System (INIS)

Croxfod, M.G.; Harwood, C.; Hall, P.C.

1992-04-01

RELAP5/MOD2 is being used by GDCD for calculation of certain small break loss-of-coolant accidents and pressurized transients in the Sizewell ''B'' PWR. To test the ability of RELAP5/MOD2 to model the primary feed-and-bleed recovery procedure following a complete loss- of-feedwater event, post test calculations have been carried out of OECD LOFT test LP-FW-01. This report describes the comparison between the code calculations and the test data. It is found that although the standard version of RELAP5/MOD2 gives a reasonable prediction of the experimental transient, the long term pressure history is better calculated with a modified code version containing a revised horizontal stratification entrainment model. The latter allows an improved calculation of entrainment of liquid from the hot leg into the surge line. RELAP5/MOD2 is found to give a more accurate simulation of the experimental transient than was achieved in previous UK studies using RETRAN-02/MOD2

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Directory of Open Access Journals (Sweden)

Eltayeb Yousif

2017-01-01

Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

International Nuclear Information System (INIS)

Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

2005-01-01

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

Developmental assessment of the SCDAP/RELAP5 code

International Nuclear Information System (INIS)

Harvego, E.A.; Slefken, L.J.; Coryell, E.W.

1997-01-01

The development and assessment of the late-phase damage progression models in the current version (designated MOD3.2) of the SCDAP/RELAP5 code are described. The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the US Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems (RCS) during severe accident conditions. Recent modeling improvements made to the MOD3.2 version of the code include (1) molten pool formation and heat up, including the transient start-up of natural circulation heat transfer, (2) in-core molten pool thermal-mechanical crust failure, (3) the melting and relocation of upper plenum structures, and (4) improvements in the modeling of lower plenum debris behavior and the potential for failure of the lower head. Finally, to eliminate abrupt transitions between core damage states and provide more realistic predictions of late phase accident progression phenomena, a transition smoothing methodology was developed and implemented that results in the calculation of a gradual transition from an intact core geometry through the different core damage states leading to molten pool formation. A wide range of experiments and modeling tools were used to assess the capabilities of MOD3.2. The results of the SCDAP/RELAP5/MOD3.2 assessment indicate that modeling improvements have significantly enhanced the code capabilities and performance in several areas compared to the earlier code version. New models for transition smoothing between core damage states, and modeling improvements/additions for cladding oxide failure, molten pool behavior, and molten pool crust failure have significantly improved the code usability for a wide range of applications and have significantly improved the prediction of hydrogen production, molten pool melt mass and core melt relocation time

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

Harvego, E. A.; Siefken, L. J.

2000-01-01

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 fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

Assessment of the RELAP5 multi-dimensional component model using data from LOFT test L2-5

International Nuclear Information System (INIS)

Davis, C.B.

1998-01-01

The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the LOFT L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident. Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L2-5 experiment were performed using the RELAP5-3D Version BF02 computer code. The calculated thermal-hydraulic responses of the LOFT primary and secondary coolant systems were generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP/MOD3

Utilization of Relap 5 computer code for analyzing thermohydraulic projects

International Nuclear Information System (INIS)

Silva Filho, E.

1987-01-01

This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt

RELAP5/MOD3.2 investigation of loss of in-house supply power for WWER 1000/320V

International Nuclear Information System (INIS)

Gencheva, R.; Pavlova, M.; Groudev, P.

2001-01-01

This paper discusses the results of the thermal-hydraulic investigations of the 'Loss of in-house supply power' accident at the Kozloduy NPP Unit 6. The RELAP5/MOD3.2 computer code has been used to stimulate the loss of in-house supply power accident in a WWER 1000 Nuclear Power Plant model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The investigation of 'Loss of normal and reverse AC power' is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the WWER 1000 against experimental transient data obtained from Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement

RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best estimate transient 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, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

RELAP5: Applications to high fidelity simulation

International Nuclear Information System (INIS)

Johnsen, G.W.; Chen, Y.S.

1988-01-01

RELAP5 is a pressurized water reactor system transient simulation code for use in nuclear power plant safety analysis. The latest version, MOD2, may be used to simulate and study a wide variety of abnormal events, including loss-of-coolant accidents, operational transients, and transients in which the entire secondary system must be modeled. In this paper, a basic overview of the code is given, its assessment and application illustrated, and progress toward its use as a high fidelity simulator described. 7 refs., 7 figs

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

SCDAP/RELAP5/MOD2 code manual

International Nuclear Information System (INIS)

Allison, C.M.; Johnson, E.C.

1989-09-01

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, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code

RELAP5-3D User Problems

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan

2002-09-01

The Reactor Excursion and Leak Analysis Program with 3D capability1 (RELAP5-3D) is a reactor system analysis code that has been developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for the U. S. Department of Energy (DOE). The 3D capability in RELAP5-3D includes 3D hydrodynamics2 and 3D neutron kinetics3,4. Assessment, verification, and validation of the 3D capability in RELAP5-3D is discussed in the literature5,6,7,8,9,10. Additional assessment, verification, and validation of the 3D capability of RELAP5-3D will be presented in other papers in this users seminar. As with any software, user problems occur. User problems usually fall into the categories of input processing failure, code execution failure, restart/renodalization failure, unphysical result, and installation. This presentation will discuss some of the more generic user problems that have been reported on RELAP5-3D as well as their resolution.

LFTPLT8: plotter program for RELAP5 code

International Nuclear Information System (INIS)

Yamano, Kazuaki; Abe, Nobuaki; Tasaka, Kanji

1981-02-01

The plotter program LFTPLT8 is a new version of the LFTPLT7 developed to plot the calculated results by RELAP5 code. The RELAP5/MOD0 code has also been revised for LFTPLT8. LFTPLT8 is capable of multiple plotting of any combination of experimental data and calculated results by RELAP4J, RELAP4/MOD5, ALARM-P1, and RELAP5/MOD0. (author)

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

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 accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

RELAP5-3D User Problems

International Nuclear Information System (INIS)

Riemke, Richard Allan

2001-01-01

The Reactor Excursion and Leak Analysis Program with 3D capability (RELAP5-3D) is a reactor system analysis code that has been developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for the U. S. Department of Energy (DOE). The 3D capability in RELAP5-3D includes 3D hydrodynamics and 3D neutron kinetics. Assessment, verification, and validation of the 3D capability in RELAP5-3D is discussed in the literature. Additional assessment, verification, and validation of the 3D capability of RELAP5-3D will be presented in other papers in this users seminar. As with any software, user problems occur. User problems usually fall into the categories of input processing failure, code execution failure, restart/renodalization failure, unphysical result, and installation. This presentation will discuss some of the more generic user problems that have been reported on RELAP5-3D as well as their resolution

RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2018-01-01

SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar. Kata kunci

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

Relap5 simulation for severe accident analysis of RSG-GAS Reactor

International Nuclear Information System (INIS)

Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

2018-01-01

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

SCDAP/RELAP5/MOD2 code manual

International Nuclear Information System (INIS)

Allison, C.M.; Johnson, E.C.

1989-09-01

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, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

Energy Technology Data Exchange (ETDEWEB)

Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Anders, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC) system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.

Methodology, status, and plans for development and assessment of the RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.W.; Riemke, R.A. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

TMI-2 analysis using SCDAP/RELAP5/MOD3.1

International Nuclear Information System (INIS)

Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M.; Dobbe, C.A.

1994-11-01

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 predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. 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 accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

International Nuclear Information System (INIS)

Meng Lin; Dong Hou; Zhihong Xu; Yanhua Yang; Ronghua Zhang

2006-01-01

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 the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, just can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is

RELAP5/MOD2 code assessment

International Nuclear Information System (INIS)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

RELAP5/MOD2 code assessment

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-11-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

RELAP5 analysis of PKL, main steam line break test

Energy Technology Data Exchange (ETDEWEB)

Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

2013-12-15

Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

ATLAS Cold Leg Top Slot Break Analysis using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Sung, Haejung; Lee, Sang Ik; Park, Ju-Hyun; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission (US-NRC) has been reviewing the design certification application for APR1400 submitted by Korea Electric Power Corporation (KEPCO). The main concern about cold leg top slot break is that cladding temperature might be increased by core uncover due to four loop seal reformation following flooding of safety injection water. An integral effect test for cold leg top slot break was performed by KAERI (Korea Atomic Energy Research Institute) using ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which is a scaled down experimental facility for APR1400. In this study, RELAP5/MOD3.3/Patch04 is assessed by experimental result of ATLAS cold leg top slot break. Also, thermal hydraulic phenomena by four loop seals reformation is observed by RELAP5 result. The RELAP5/MOD3.3/Patch04 is assessed by the experimental result of ATLAS cold leg top slot break. The top slot break is described by offtake model, and the mass flow rate is fairly well estimated. The RELAP5 well predicts the correlation between general trend and four loop seal reformation. The pressure of the core region and the cladding temperature tends to increase during four loop seal reformation due to steam path blockage on four loop seals. It is presumed that the code cannot estimate two phase phenomena by loop seal clearing as same as experiments. In terms of cladding temperature, loop seal reformation due to loop seal elevation of APR1400 does not need to be the issue, since the void fraction at the active top core is maintained over 0.4.

Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior

International Nuclear Information System (INIS)

Allison, C.M.; Rempe, J.L.; Chavez, S.A.

1994-11-01

The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and in combination with VICTORIA, fission product release and transport during severe accidents. Improvements for existing debris bed and molten pool models in the SCDAP/RELAP5/MOD3.1 code are described in this report. Model improvements to address (a) debris bed formation, heating, and melting; (b) molten pool formation and growth; and (c) molten pool crust failure are discussed. Relevant data, existing models, proposed modeling changes, and the anticipated impact of the changes are discussed. Recommendations for the assessment of improved models are provided

Analysis Of Control Rod Ejection Of APR1400 By RELAP5

International Nuclear Information System (INIS)

Le Thi Thu; Hoang Minh Giang; Vo Thi Huong; Le Dai Dien

2011-01-01

This paper presents the analysis of Reactivity Induced Accident caused by ejection of a Control Element Assembly (CEA) from APR 1400 reactor vessel within 0.05 second. The initial condition were assumed as following: power level at 102%, delayed neutron fraction β = 412 pcm and CEA worth = 110 pcm. The analysis was simulated by RELAP5 code through two step: calculation of steady state and calculation of transient with initial condition mentioned as above. Some output results were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as power, reactivity feedback from fuel temperature changes (Doppler) as well as temperature, pressure, DNBR within 6 second of the accident. (author)

RELAP5/MOD3 code coupling model

International Nuclear Information System (INIS)

Martin, R.P.; Johnsen, G.W.

1994-01-01

A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability

PACTEL ISP-33. RELAP5 assessment

International Nuclear Information System (INIS)

Siccama, N.B.; Roodbergen, H.A.

1995-08-01

This report presents the results of the calculation of the PACTEL ISP-33 experiment as obtained from the RELAP5 code. The main goal of the ECN contribution to this ISP was to assess RELAP5 on one- and two-phase natural circulation phenomena which occur in Eastern European VVER plants in case of LOCA conditions. Different natural circulation modes were calculated in the simulation of the ISP-33 experiment. The single-phase liquid flow, the steady two-phase flow, and the boiler-condenser single-phase heat removal are calculated well by the RELAP5 code. The phenomena in the transient two-phase flow are difficult to simulate. (orig.)

RHF RELAP5 model and preliminary loss-of-offsite-power simulation results for LEU conversion

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

2014-08-01

The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

RELAP5 modification for CHEOPE simulations

International Nuclear Information System (INIS)

Casamirra, M.; Castiglia, F.; Giardina, M.; Meloni, P.

2005-01-01

Nowadays, an extensive R and D program is in progress in order to maximize the safety of the Lead-Bismuth Eutectic alloy (LBE) ADS plants and optimise their layout. Recently, in the framework of the collaboration between the Department of Nuclear Engineering of the University of Palermo and the ENEA ''E. Clementel'' Research Centre in Bologna, validation works on the LBE RELAP5mod3.2.2a code, were carried out. The validation was based on the experimental tests performed on the CHEOPE (CHEmical OPErational transient) facility, an experimental rig built up to evaluate the heat transfer performances between the LBE and a diathermic oil secondary fluid, in support to the MEGAPIE (MEGAwatt PIlot Experiment) project, aimed at testing the LBE spallation target technology at 1 MW proton-beam power. As expected, the comparison between the experimental results and post test calculations highlighted not fully satisfactory RELAP5 evaluations of the oil-side heat transfer coefficient, due to RELAP5 code inadequacies in predicting the 3D effects induced in the oil bulk by the facility peculiar geometry, about which is referred in the following. In the present paper we have suitably modified the RELAP5 heat transfer model, to take into account the consequent improved heat transfer conditions. The obtained results show very good performance of the so modified RELAP5 code. (author)

RELAP5 nuclear plant analyzer capabilities

International Nuclear Information System (INIS)

Wagner, R.J.; Ransom, V.H.

1982-01-01

An interactive execution capability has been developed for the RELAP5 code which permits it to be used as a Nuclear Plant Analyzer. This capability has been demonstrated using a simplified primary and secondary loop model of a PWR. A variety of loss-of-feed-water accidents have been simulated using this model. The computer execution time on a CDC Cyber 176 is one half of the transient simulation time so that the results can be displayed in real time. The results of the demonstration problems are displayed in digital form on a color schematic of the plant model using a Textronics 4027 CRT terminal. The interactive feature allows the user to enter commands in much the same manner as a reactor operator

Study on coupling of three-dimension space time neutron kinetics model and RELAP5 and improvement of RELAP5

International Nuclear Information System (INIS)

Gui Xuewen; Cai Qi; Luo Bangqi

2007-01-01

A two-group three-dimension space-time neutron kinetics model is applied to the RELAP5 code, which replaces the point reactor kinetics model. A visual operation interface is designed to convenience interactive operation between operator and computer. The calculation results and practical applications indicate that the functions and precision of improved RELAP5 are enhanced and can be easily used. The improved RELAP5 has a good application perspective in nuclear power plant simulation. (authors)

Comparision of calculations for the ROSA-IV LSTF with RELAP5/MOD0 and RELAP5/MOD1 (cycle 1)

International Nuclear Information System (INIS)

Fineman, C.P.; Tanaka, Mitsugu; Tasaka, Kanji

1982-03-01

10% and 2.5% cold leg break analyses have been completed for the ROSA-IV Large Scale Test Facility (LSTF) with the RELAP5/MOD0 and RELAP5/MOD1, cycle 1, computer codes. Comparisons between the calculations were made to determine any differences in the results obtained from the two versions of RELAP5. Differences in the two calculations were found which can be attributed to changes in the flow regime maps and critical flow model. (author)

Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

International Nuclear Information System (INIS)

Vorob'ev, Yu.; Zhabin, O.

2015-01-01

Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

RELAP/FRAP-T6 analysis of seized and sheared shaft accidents

International Nuclear Information System (INIS)

Bollinger, J.S.; Ito, T.; Peeler, G.B.

1984-01-01

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 shaft seizure event was modeled in RELAP5 by using the pump model shaft stop option. In modeling the sheared shaft failure, the faulted pump was replaced with a branch component having no flow losses. In general, the RELAP5-predicted system response for the seized shaft transient was very comparable to the results presented in the Seabrook FSAR, although the Reactor Coolant System (RCS) pressure response was somewhat different. The RELAP5 sheared-shaft analysis results were very similar to those for the seized shaft

The RELAP5-Based NPA of the VVER Type Paks NPP

International Nuclear Information System (INIS)

Guba, A.; Toth, I.; Mandy, C.; Stubbe, E.

1999-01-01

NPA is a data driven interactive graphical tool for visualisation of different plant conditions. Data generated by the analysis code RELAP5/MOD3.2 are processed and displayed on a computer monitor. The NPA model of Paks NPP Unit 3 was developed with the aim to demonstrate the phenomena occurring in different transient/accident scenarios. This VVER-specific NPA development is a result of a cooperation between BELGATOM and KFKI-AEKI. (author)

Steady-state simulations of a 30-tube once-through steam generator with the RELAP5/MOD3 and RELAP5/MOD2 computer codes

International Nuclear Information System (INIS)

Hassan, Y.A.; Salim, P.

1991-01-01

This paper reports on a steady-state analysis of a 30-tube once-through steam generator that has been performed on the RELAPS/MOD3 and RELAPS/MOD2 computer codes for 100, 75, and 65% loads. Results obtained are compared with experimental data. The RELAP5/MOD3 results for the test facility generally agree reasonably well with the data for the primary-side temperature profiles. The secondary-side temperature profile predicted by RELAP5/MOD3 at 75 and 65% loads agrees fairly well with the data and is better than the RELAP5/MOD2 results. However, the RELAP5/MOD3 calculated secondary-side temperature profile does not compare well with the 100% load data

Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

2004-01-01

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini

2011-01-01

RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)

Evaluation of RELAP5 MOD 3.1.1 code with GIRAFFE Test Facility: Phase 1, Step 2 nitrogen venting tests

International Nuclear Information System (INIS)

Boyer, B.D.; Slovik, G.C.; Rohatgl, U.S.

1995-01-01

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 noncondensable gases during an accident. To model the transient behavior of the SBWR with a system code, the code should properly simulate the expected phenomena. To validate the applicability of RELAP5 MOD 3.1.1, the data from three Phase 1, Step 2 nitrogen venting tests at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal facility and RELAP5 calculations of these tests were compared. The comparison of the GIRAFFE data against the results from the RELAP5 calculations showed that it can predict condensation and gas purging phenomena occurring in the long-term decay heat rejection phase. In this phase of the transient, condensation in the PCCS is the only means to reject heat from the SBWR containment. In the two tests where the nitrogen purge vent line was at its deepest submergence in the Suppression Pool (SIP), the RELAP5 results mirrored the behavior of the containment pressures and of the water levels in the Horizontal Vent (HV) and the nitrogen purge line tube of the GIRAFFE data. However, in the test with the shallowest purge line submergence, there was appreciable direct contact condensation on the pool surface of the HV despite modeling efforts to deter these phenomena. This surface condensation, unobserved in the GIRAFFE tests, was a major cause of RELAP5 predicting early containment depressurization and the subsequent early rise in HV and nitrogen purge line water levels. The present RELAP5 MOD3.1.1 interfacial heat and mass transfer model does not properly degrade direct contact steam condensation in the presence of noncondensable gases sitting on a pool

Streamlining of the RELAP5-3D Code

International Nuclear Information System (INIS)

Mesina, George L; Hykes, Joshua; Guillen, Donna Post

2007-01-01

RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements have been made decrease code run time and increase solution accuracy. These include vectorization, parallelization, use of improved equation solvers for thermal hydraulics and neutron kinetics, and incorporation of improved library utilities. In the area of applied nuclear engineering, expanded capabilities include boron and level tracking models, radiation/conduction enclosure model, feedwater heater and compressor components, fluids and corresponding correlations for modeling Generation IV reactor designs, and coupling to computational fluid dynamics solvers. Ongoing and proposed future developments include improvements to the two-phase pump model, conversion to FORTRAN 90, and coupling to more computer programs. This paper summarizes the general improvements made to RELAP5-3D, with an emphasis on streamlining the code infrastructure for improved maintenance and development. With all these past, present and planned developments, it is necessary to modify the code infrastructure to incorporate modifications in a consistent and maintainable manner. Modifying a complex code such as RELAP5-3D to incorporate new models, upgrade numerics, and optimize existing code becomes more difficult as the code grows larger. The difficulty of this as well as the chance of introducing errors is significantly reduced when the code is structured. To streamline the code into a structured program, a commercial restructuring tool, FOR( ) STRUCT, was applied to the RELAP5-3D source files. The

RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-08-15

Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

International Nuclear Information System (INIS)

Blanchat, T.; Hassan, Y.

1989-01-01

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

Energy Technology Data Exchange (ETDEWEB)

Protze, M. [Siemens-KWU, Erlangen (Germany)

1995-12-31

Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

Energy Technology Data Exchange (ETDEWEB)

Protze, M [Siemens-KWU, Erlangen (Germany)

1996-12-31

Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

Visualization of RELAP5-3D best estimate code

International Nuclear Information System (INIS)

Mesina, G.L.

2004-01-01

The Idaho National Engineering Laboratory has developed a number of nuclear plant analysis codes such as RELAP5-3D, SCDAP/RELAP5-3D, and FLUENT/RELAP5-3D that have multi-dimensional modeling capability. The output of these codes is very difficult to analyze without the aid of visualization tools. The RELAP5-3D Graphical User Interface (RGUI) displays these calculations on plant images, functional diagrams, graphs, and by other means. These representations of the data enhance the analysts' ability to recognize plant behavior visually and reduce the difficulty of analyzing complex three-dimensional models. This paper describes the Graphical User Interface system for the RELAP5-3D suite of Best Estimate codes. The uses of the Graphical User Interface are illustrated. Examples of user problems solved by use of this interface are given. (author)

Development of a system for a linked analysis of RELAP5 and MAAP4 and its application for an analysis of LB-LOCA case of APR1400

International Nuclear Information System (INIS)

Park, Chang Hwan; Lee, Un Chul; Park, Goon Cherl; Suh, Kune Yull

2004-01-01

The RELAP5 (1995) and MAAP4 (1994) linked-analysis system was designed and illustrative calculation was performed. A large-break loss-of-coolant accident (LBLOCA) was taken as the reference case for the APR1400 (Advanced Power Reactor 1400MWe). For the early phase of this case, the calculational results of two codes have some deviations in water level depletion and hot assembly temperature. Ordinarily, it is considered that RELAP5 has enough accuracy in calculation of the thermal hydraulic behavior of typical PWR during design basis accidents. If the data set for the thermal hydraulic state of RELAP5 can be well-transferred to MAAP4 as an initial condition, the overall transients given by the linked analysis can get more reliability. In this study, the linked analysis system of RELAP5 and MAAP4 doesn't mean the mechanically integrated code structure. The objective of this study is to formulate the linked analysis system of RELAP5 and MAAP4, which should precede construction of mechanically integrated analyzer. Thus, the main scope of this work covers development of the methodology for data linkage and decision of the transfer timing

Modeling a Printed Circuit Heat Exchanger with RELAP5-3D for the Next Generation Nuclear Plant

International Nuclear Information System (INIS)

2010-01-01

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 simulations. For purposes of simplicity, a straight channel configuration was assumed. A parallel intermediate heat exchanger configuration was assumed for the RELAP5 model design. The RELAP5 modeling also required the semicircular channels in the heat exchanger to be mapped to rectangular channels. The initial RELAP5 run outputs steady state conditions which were then compared to the heat exchanger performance theory to ensure accurate design is being simulated. An exponential loss of pressure transient was simulated. This LOCA describes a loss of coolant pressure in the primary side over a 20 second time period. The results for the simulation indicate that heat is initially transferred from the primary loop to the secondary loop, but after the loss of pressure occurs, heat transfers from the secondary loop to the primary loop.

Modification of the bubble rise model used in RELAP4/Mod5 computer code for transients analysis

International Nuclear Information System (INIS)

Scharfmann, E.

1981-01-01

To improve the separation phase and heat transfer models in RELAP4/MOD5 computer code, in order to make more realistic estimates of the thermohydraulic behavior of the core submitted to a loss-of-coolant accident, is the objective of this work. This research is directed to the accident analysis caused by small breaks in the primary circuit of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of RELAP code, and the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

Analysis of eventual accidents in a water experimental loop, using the Relap 4 computer code

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1981-01-01

Transients caused by accidents as (1) loss of coolant, (2) failure in the principal pump and (3) power excursions were analysed. In the accident simulation, the Relap 4/Mod 3 computer code was used. The results obtained with the steady state model showed to be consistent with the project-and operation data of the experimental loop. For all the accidents analysed that considered the performance of safety systems, the highest temperature of the heating rods in the testing section did not exceed the permissible temperature. (E.G.) [pt

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

Directory of Open Access Journals (Sweden)

Itamar Iliuk

2016-01-01

Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

An implicit steady-state initialization package for the RELAP5 computer code

International Nuclear Information System (INIS)

Paulsen, M.P.; Peterson, C.E.; Odar, F.

1995-08-01

A direct steady-state initialization (DSSI) method has been developed and implemented in the RELAP5 hydrodynamic analysis program. It provides a means for users to specify a small set of initial conditions which are then propagated through the remainder of the system. The DSSI scheme utilizes the steady-state form of the RELAP5 balance equations for nonequilibrium two-phase flow. It also employs the RELAP5 component models and constitutive model packages for wall-to-phase and interphase momentum and heat exchange. A fully implicit solution of the linearized hydrodynamic equations is implemented. An implicit coupling scheme is used to augment the standard steady-state heat conduction solution for steam generator use. It solves the primary-side tube region energy equations, heat conduction equations, wall heat flux boundary conditions, and overall energy balance equation as a coupled system of equations and improves convergence. The DSSI method for initializing RELAP5 problems to steady-state conditions has been compared with the transient solution scheme using a suite of test problems including; adiabatic single-phase liquid and vapor flow through channels with and without healing and area changes; a heated two-phase test bundle representative of BWR core conditions; and a single-loop PWR model

RELAP5 simulation of a large break Loss of Coolant Accident (LOCA) in the hot leg of the primary system in Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Andrade, Delvonei Alves de; Sabundjian, Gaiane

2004-01-01

The objective of this work is to present the simulation of a large break loss of coolant accident - LBLOCA in the hot leg of the primary loop in Angra 2, with RELAP5/MOD3.2.2g code. This accident is described in the Final Safety Report Analysis of Angra 2 - FSAR and consists basically of the hot leg total break, in loop 20 of the plant. The area considered for the rupture is 4480 cm 2 , which corresponds to 100% of the pipe flow area. Besides, this work also has the objective of verifying the efficiency of the emergency core coolant system - ECCS in case of accidents and transients. The thermal-hydraulic processes inherent to the accident phenomenology, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the liquid level, until the ECCS is capable to reflood it

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

International Nuclear Information System (INIS)

Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F.; Sabundjian, Gaiane

2013-01-01

To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

Energy Technology Data Exchange (ETDEWEB)

Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F., E-mail: plobo.a@uol.com.br, E-mail: fbraz@ieav.cta.br, E-mail: eduardo@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

Energy Technology Data Exchange (ETDEWEB)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

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 HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

International Nuclear Information System (INIS)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

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 HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs

Development and assessment of the COBRA/RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

1997-04-01

The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

International Nuclear Information System (INIS)

Xie Heng

2017-01-01

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA. (author)

Post-test analysis of LOBI BT-01 using RELAP5/MOD2 and RELAP5/MOD3

International Nuclear Information System (INIS)

Holmes, B.J.

1991-08-01

LOBI is a high pressure, electrically heated integral system test facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. This report describes the results of an analysis of test BT-01, which simulates a 10% steam line break. The bulk of the analysis was performed using the Project Version of RELAP5/MOD2, with additional calculations using RELAP5/MOD3 for comparison. The codes provided generally good agreement with data. In particular, the break flows were well modelled, although the mass flow data proved to be unreliable, and this conclusion had to be derived from interpreting other signals. RELAP over-predicted primary/secondary heat transfer in the broken loop, however, leading to a more rapid cool-down of the primary circuit. Furthermore, the primary side pressure response was critically dependent upon the pressuriser behaviour, and the correct timing of the uncovery of the surge line. Inter-phase drag was not well predicted in the broken loop steam generator intermals, although some improvement was seen in the RELAP5/MOD3 predictions. MOD3 gave a reduction in primary/secondary heat transfer during the test pre-conditioning phase, resulting in a lower secondary side pressure at the start of the transient compared with MOD2. (author)

RELAP5/MOD3 code manual: User's guide and input requirements. Volume 2

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best estimate transient 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, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation

Development of the unified version of COBRA/RELAP5

Energy Technology Data Exchange (ETDEWEB)

Jeong, J J; Ha, K S; Chung, B D; Lee, W J; Sim, S K [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

The COBRA/RELAP5 code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERI developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan. 13 refs., 5 figs., 2 tabs. (Author)

Development of the unified version of COBRA/RELAP5

Energy Technology Data Exchange (ETDEWEB)

Jeong, J. J.; Ha, K. S.; Chung, B. D.; Lee, W. J.; Sim, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

The COBRA/RELAP5 code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERI developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan. 13 refs., 5 figs., 2 tabs. (Author)

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

International Nuclear Information System (INIS)

Bencik, V.; Cavlina, N.; Grgic, D.

2012-01-01

The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

Evaluation and assessment of reflooding models in RELAP5/Mod2.5 and RELAP5/Mod3 codes using Lehigh University and PSI-Neptun bundle experimental data

Energy Technology Data Exchange (ETDEWEB)

Sencar, M.; Aksan, N. [Paul Scherrer Institute, Villigen (Switzerland)

1995-09-01

An extensive analysis and assessment work on reflooding models of RELAP5/Mod2.5 and, RELAP5/Mod3/v5m5 and RELAP/Mod3/v7j have been performed. Experimental data from LehighUniversityv. and PSI-NEPTUN bundle reflooding experiments have been used for the assessment, since both of these tests cover a broad range of initial conditions. Within the range of these initial conditions, it was tried to identify their separate impacts on the calculated results. A total of six Lehigh University reflooding bundle tests and two PSI-NEPTUN tests with bounding initial conditions are selected for the analysis. Detailed nodalisation studies both for hydraulic and conduction heat transfer were done. On the basis of the results obtained from these cases, a base nodalisation scheme was established. All the other analysis work was performed by using this base nodalisation. RELAP5/Mod2.5 results do not change with renodalisation but RELAP5/Mod3 results are more sensitive to renodalisation. The results of RELAP5/Mod2.5 versions show very large deviations from the used experimental data. These results indicate that some of the phenomenology of the events occurring during the reflooding could not be identified. In the paper, detailed discussions on the main reasons of the deviations from the experimental data will be presented. Since, the results and findings of this study are meant to be a developmental aid, some recommendations have been drawn and some of these have already been implemented at PSI with promising results.

Application of RELAP/SCDAPSIM to TRIGA system thermal hydraulic analysis

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.; Quamrul Huda, Md.

2010-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. This paper describes the development of a representative input model for the 3MW TRIGA research reactor at AERE Bangladesh, describes the testing and qualification of the model using MOD4.0 advanced input checking and graphical display options, and then presents representative results for selected calculations. (author)

RELAP5/MOD3 AP600 problems

International Nuclear Information System (INIS)

Riemke, R.A.

1993-01-01

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 Assessment and Applications Program (ICAP). The code is currently being used to simulate transients for the next generation of advanced light water reactors (ALWR's). One particular reactor design is the Westinghouse AP600 pressurized water reactor (PWR), which consists of two hot legs and four cold legs as well as passive emergency core cooling (ECC) systems. Initial calculations with RELAP5/MOD3 indicated that the code was not as robust as RELAP5/MOD2.5 with regard to AP600 calculations. Recent modifications in the areas of condensation wall heat transfer, interfacial heat transfer in the presence of noncondensibles, bubbly flow interfacial heat transfer, and time smoothing of both interfacial drag and interfacial heat transfer have improved the robustness, although more reliability is needed

RELAP5 analysis of PACTEL injection tests

International Nuclear Information System (INIS)

Kimber, G.R.; Lillington, J.N.

2000-01-01

A characteristic feature of advanced reactor designs is their reliance on passive safety systems. It is important to assess both the operation of such systems and the ability of systems codes, such as RELAP5, to model them. In Finland VTT Energy, together with Lappeenranta University of Technology, is using the PACTEL facility for the investigation of passive core cooling systems. In particular, a core make-up tank (CMT) has been installed in the rig to operate in a similar manner to those in many Advanced PWR designs. Three small break tests, GDE-24, GDE-34 and GDE-43 in the PACTEL facility were chosen for modelling with RELAP5. The objective of GDE-24 was to investigate CMT behaviour and in particular the effects of condensation in the CMT. The second test, GDE-34, was similar except that it had a smaller CMT and at the start of the test the water in the CMT and connecting pipework was at an elevated temperature. Test GDE-43 focused on conditions when the driving force for flow through the passive system injection system (PSIS) slowly disappears. Analysis of all tests reported here was carried out with RELAP5/MOD 3.2.1.2. The paper summarises the conclusions of all the tests. A critical part of the study revolved around modelling of the CMT. A model was developed to allow its detailed behaviour to be investigated more easily. This enabled recommendations for improving the condensation modelling in RELAP5 to be made. Apart from the wall condensation modelling issue, the implication of the work is that RELAP5/MOD 3.2.1.2 (a comparatively recent version of the code) is broadly adequate for these applications. (author)

A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Andreani, M. [Swiss Federal Institute of Technology, Zurich (Switzerland); Analytis, G.T.; Aksan, S.N. [Paul Scherrer Institute, Villigen (Switzerland)

1995-09-01

The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.

RELAP5/MOD1-EUR evaluation. Comparison with the INEL original version

International Nuclear Information System (INIS)

Mazzantini, O.A.

1990-01-01

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 improved by EURATOM (RELAP5/MOD1-EUR) which was transferred to ENACE through agreements made with SIEMENS/KWU. (Author) [es

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

International Nuclear Information System (INIS)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C.; Rohatgi, U.S.

1995-01-01

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 System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations al these conditions were compared with the GIRAFFE data. The effects of PCCS cell nodings on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three-node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer in the presence of noncondensable gases with only a coarse mesh. The cell length term in the condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

International Nuclear Information System (INIS)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C.

1995-01-01

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 System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

Energy Technology Data Exchange (ETDEWEB)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C. [and others

1995-09-01

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 System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {plus_minus}5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang Jinzhao

2004-01-01

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) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

Coupling of RELAP5-3D and GAMMA codes for Nuclear Hydrogen System Analysis

International Nuclear Information System (INIS)

Jin, Hyung Gon

2007-02-01

RELAP5-3D is one of the most important system analysis codes in nuclear field, which has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The GAMMA code is a multi-dimensional multi-component mixture analysis code with the complete set of chemical reaction models which is developed for safety analysis of HTGR (High Temperature Gas Cooled Reactor) air-ingress. The two codes, RELAP5-3D and GAMMA, are coupled to be used for nuclear-hydrogen system analysis, which requires the capability of the analysis of multi-component gas mixture and two-phase flow. In order to couple the two codes, 4 steps are needed. Before coupling, the GAMMA code was transformed into DLL (dynamic link liberally) from executive type and RELAP5-3D was recompiled into Compaq Visual Fortran environments for our debugging purpose. As the second step, two programs - RELAP5-3D and GAMMA codes - must be synchronized in terms of time and time step. Based on that time coupling, the coupled code can calculate simultaneously. Time-step coupling had been accomplished successfully and it is tested by using a simple test input. As a next step, source-term coupling was done and it was also tested in two different test inputs. The fist case is a simple test condition, which has no chemical reaction. And the other test set is the chemical reaction model, including four non-condensable gas species, which are He, O2, CO, CO2. Finally, in order to analyze combined cycle system, heat-flux coupling has been made and a simple heat exchanger model was demonstrated

Break model comparison in different RELAP5 versions

International Nuclear Information System (INIS)

Parzer, I.

2003-01-01

The presented work focuses on the break flow prediction in RELAP5/MOD3 code, which is crucial to predict core uncovering and heatup during the Small Break Loss-of-Coolant Accidents (SB LOCA). The code prediction has been compared to the IAEA-SPE-4 experiments conducted on the PMK-2 integral test facilities in Hungary. The simulations have been performed with MOD3.2.2 Beta, MOD3.2.2 Gamma, MOD3.3 Beta and MOD3.3 frozen code version. In the present work we have compared the Ransom-Trapp and Henry-Fauske break model predictions. Additionally, both model predictions have been compared to itself, when used as the main modeling tool or when used as another code option, as so-called 'secret developmental options' on input card no.1. (author)

RELAP5-3D Code Includes ATHENA Features and Models

International Nuclear Information System (INIS)

Riemke, Richard A.; Davis, Cliff B.; Schultz, Richard R.

2006-01-01

Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF 6 , xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)

Application of RELAP5 to a pipe blowdown experiment

International Nuclear Information System (INIS)

Carlson, K.E.; Ransom, V.H.; Wagner, R.J.

1980-01-01

The application of the RELAP5 computer program to a pipe blowdown experiment is described in this paper. The basic hydrodynamic model, constitutive relations, and special process models included in RELAP5 are also briefly discussed. The results of this application confirm the effectiveness of using a choked flow model

Uncertainty in RELAP5/MOD3.2 calculations for interfacial drag in downward two-phase flow

International Nuclear Information System (INIS)

Clark, Collin; Schlegel, Joshua P.; Hibiki, Takashi; Ishii, Mamoru; Kinoshita, Ikuo

2016-01-01

Highlights: • Uncertainty propagation is key for best estimate code reliability. • Uncertainty in drift flux correlations used to evaluate uncertainty in interfacial drag. • Bias and error have been compared for various models. - Abstract: RELAP5/MOD3.2 is a thermal-hydraulic system analysis code used to predict the response of nuclear reactor coolant systems in the event of certain accident scenarios. It is important that RELAP and other system analysis codes are able to accurately predict various two-phase flow phenomena, particularly the interfacial transfers between the liquid and gas phases. It is also important to understand how much uncertainty exists in these predictions due to uncertainties in the constitutive relations used to close the two-fluid model. In this paper, the uncertainty in the interfacial drag calculated by RELAP5/MOD3.2 due to errors in the drift-flux models used to close the model is evaluated and compared to the correlation developed by Goda et al. (2003). The case of downward flow is considered due to the importance of co-current and counter-current downward flow for predicting behavior in the downcomer of reactor systems during small-break Loss of Coolant Accidents (LOCAs) in nuclear reactor systems. The overall uncertainty in the interfacial force calculations due to error in the distribution parameter models were found to have a bias of +8.1% and error of 20.1% for the models used in RELAP5, and a bias of −30.8% and error of 23.1% for the correlation of Goda et al. (2003). However this analysis neglects the effects of compensating errors in the drift-flux parameters, as the drift velocity is assumed to be perfectly accurate. More physically meaningful results could be obtained if the distribution parameter and drift velocity were calculated directly from local phase concentration and velocity measurements, however no studies were available which included all of this information.

Comparative severe accident analysis of WWER 1000/B 320 LOCA DN100 computed by computer codes ASTEC V1.1 and SCDAP/RELAP5

International Nuclear Information System (INIS)

Kalchev, B.; Dimov, D.; Tusheva, P.; Mladenov, I.

2005-01-01

This paper presents the modelling approach for LOCA 100 mm sequence for WWER 1000-B 320 type of reactor with the integral ASTEC computer code and SCDAP/RELAP5 computer code. As a basic input deck the reference input file for Balakovo NPP from the released ASTEC CD has been applied. As a first part of the calculations for the SBLOCA sequence the ASTEC v1.1 modules CESAR, DIVA and CPA have been activated in a coupled mode. For SCDAP/RELAP5 calculation input deck for WWER 1000-B 320 has been applied which meant to be closer to the initial boundary conditions applied for ASTEC WWER 1000 input deck. A SBLOCA 100 mm comparison between ASTEC v1.1 and SCADAP/RELAP5 has been presented. ASTEC predicts vessel failure at 15620 s. ASTEC and SCDAP/RELAP5 give close but not similar results - this could be observed on the trends. The comparison of 100 mm-break shows that SCDAP/RELAP5 predicts clear phenomenological changes in primary pressure evolution and molten pool formation. Similar hydrogen production mass for both codes around 5000 s is detected

Development and Assessment of the Appendix K Version of RELAP5-3D for LOCA Licensing Analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chang, C.-J.; Hung, H.-J.

2002-01-01

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. Instead, implementation of evaluation models required by Appendix K of 10CFR50 on an advanced thermal-hydraulic platform such as RELAP5, TRAC, etc., also can gain significant margin for the PCT calculation. Through compliance evaluation against Appendix K of 10CFR50, all of the required evaluation models have been implemented in RELAP5-3D. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effects experiments and eight sets of LOCA integral experiments were adopted. Through the assessments against separate-effects experiments, the success of the code modification in accordance with Appendix K of 10CFR50 was demonstrated. Besides, one set of a typical integral large-break LOCA from Loss-of-Fluid Test Facility experiments (L2-5) has also been applied to preliminarily evaluate the integral performance of the Appendix K version of RELAP5-3D. The PCT predicted by the evaluation models is greater than the one from best-estimate calculation in the whole LOCA history with the conservatism of 150 K, and the measured PCTs of L2-5 are also well bounded by the evaluation model calculation. Another seven sets of integral-effect experiments will be further applied in the next step to ensure the reasonable integral conservatism of the newly developed LOCA licensing analysis code (RELAP5-3DK/INER), which can cover all the phases of both large- and small LOCA in one code

MSLB coupled 3D neutronics-thermalhydraulic analysis of a large PWR using RELAP5-3D

International Nuclear Information System (INIS)

Lo Nigro, A.; Spadoni, A.; D'Auria, F.; Saiu, G.

2001-01-01

A RELAP5-3D model of the Westinghouse AP1000 NSSS has been set up and it has been used to analyze the MSLB accident. Main results (both spatial distributions and time trends) have been represented with 3D plots and graphical movies. The method applied allows accounting for the coupled 3D neutronics and thermalyhdraulics effects, suggesting to consider its applicability in Safety Analysis.(author)

Developmental assessment of RELAP5/MOD3 using the semiscale natural circulation tests

International Nuclear Information System (INIS)

Carlson, K.E.

1990-01-01

A code development effort creating RELAP5/MOD3 from RELAP5/MOD2 has been completed. Upon completion, a developmental assessment task was performed. One of the problems used for the developmental assessment was the Semiscale Natural Circulation Test. Calculated results from RELAP5/MOD3 are compared to measured data and previously calculated results from RELAP5/MOD2. 10 refs., 6 figs., 1 tab

Analysis of Seven NEPTUN-III (Tight-Lattice) Bottom-Flooding Experiments with RELAP5/MOD3.3/BETA

International Nuclear Information System (INIS)

Analytis, G.Th.

2004-01-01

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 a tight hexagonal lattice (which was one of the core geometries considered at the time for an HPLWR) following a hypothetical loss-of-coolant accident scenario. Even though the latest version of the code has as a default the new PSI reflood model developed by the author, which was tested and assessed against reflooding data obtained at standard light water reactor lattices, this work shows that for tight lattices, the code underpredicts the peak clad temperatures measured during a series of reflooding experiments performed at the NEPTUN-III tight-lattice heater rod bundle facility. The reasons for these differences are discussed, and the (possible) changes needed in the framework of RELAP5/MOD3.3 for improving the modeling of reflooding in tight lattices are investigated

RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation

Directory of Open Access Journals (Sweden)

Andrej Prošek

2018-01-01

Full Text Available The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Role of RELAP/SCDAPSIM in Nuclear Safety

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS) is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of some of the more important activities including the analysis of research reactors and Nuclear Power Plants, design and analysis of experiments, and training

Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

Energy Technology Data Exchange (ETDEWEB)

Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima, E-mail: dayanefs@ipen.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-07-01

The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

International Nuclear Information System (INIS)

Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima

2017-01-01

The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-10-15

Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

Modifications of the bubble rise model and heat transfer model used in RELAP 4/Mod 5 computer code for transient analysis

International Nuclear Information System (INIS)

Scharfmann, E.; Silva, D.E. da

1981-01-01

The modifications on the phase separation model and heat tranfer model in Relap4/Mod 5 computer code, in order to make more realistic estimates of the core thermohydraulic behavior submitted to a loss of coolant accident. This research is directed to the accident analysis caused by small breaks in the primary circuits of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of Relap code, as well as the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

Using computer program RELAP5/MOD2 on microcomputers

International Nuclear Information System (INIS)

Grgic, D.; Bajs, T; Cavlina, N.; Debrecin, N.

1990-01-01

Our work on installation of RELAP5/MOD2 code on IBM4341, mVAX 11, MGT-386 and COMPAQ-386/20e computers is described. Main characteristics of RELAP5/MOD2 structure programming style and differences between FORTRAN VS, VAX-11 FORTRAN and NDP FORTRAN 386 are presented. We discussed basic philosophy used in modification and testing and test results. (author)

Countercurrent flow limitation model for RELAP5/MOD3

International Nuclear Information System (INIS)

Riemke, R.A.

1991-01-01

This paper reports on a countercurrent flow limitation model incorporated into the RELAP5/MOD3 system transient analysis code. The model is implemented in a manner similar to the RELAP5 chocking model. Simulations using air/water flooding test problem demonstrate the ability of the code to significantly improve its comparison to data when a flooding correlation is used

RELAP5/MOD2 implementation on various mainframes including the IBM and SX-2 supercomputer

International Nuclear Information System (INIS)

DeForest, D.L.; Hassan, Y.A.

1987-01-01

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 Regulatory Commission (NRC). The major enhancement from RELAP5/MOD1 is the use of a six-equation, two-fluid nonequilibrium and nonhomogeneous model. Other improvements include the addition of a noncondensible gas component and the revision and addition of drag formulation, wall friction, and wall heat transfer. Several test cases were run to benchmark the IBM and SX-2 installations against the CDC computer and the CRAY-2 and CRAY/XMP. These included the Edward's pipe blow-down and two separate reflood cases developed to simulate the FLECHT-SEASET reflood test 31504 and a postcritical heat flux (CHF) test performed at Lehigh University

Investigation of the Phebus FPT0 bundle degradation with SCDAP/RELAP5

International Nuclear Information System (INIS)

Smit, S.O.; Sengpiel, W.; Hering, W.

1998-04-01

The in-pile experiment Phebus FPT0 provides an excellent data base reflecting the course and the consequences of a severe core melt accident starting from the core uncovery up to bundle degradation and molten pool formation. In the IRS post-test calculations of the Phebus FPT0 have been performed with SCDAP/RELAP5. A detailed parameter study has shown that some models used in the code still have to be improved and that some parametric models need to be substituted by more physical models. In the context of this parameter study, the heat transfer through the Phebus FPT0 shroud has been identified to be one of the most influential physical processes on the course of bundle degradation. Especially the gap behaviour and the heat transport through the gaps of the FPT0 shroud have shown to be insufficiently modeled by the original code version. Therefore, the shroud heat transfer model has been improved to consider dynamic gap closure by thermal expansion of the shroud materials and to take into account radiation heat transfer through open gaps. In this report, the results of the parameter study for FPT0 obtained with the original code are compared to the results of a reference calculation which includes the improved shroud model. It is shown that SCDAP/RELAP5 is now able to calculate the heat losses through a shroud containing gas-filled gaps like that of Phebus FPT0 quite accurately. Thus, SCDAP/RELAP5 now can also be used more successfully for test analyses of experiments like Phebus FPT1 and FPT2, and of the QUENCH test series. (orig./MM) [de

SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

International Nuclear Information System (INIS)

Davis, K.L.

1995-06-01

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 accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made

SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Davis, K.L. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

1995-06-01

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 accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made.

Assessment and improvement of condensation models in RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

Assessment and improvement of condensation models in RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

Post test analysis of TEPSS tests -P2-, -P3-, -P5- and -P7- using the system code RELAP5/MOD 3.2

International Nuclear Information System (INIS)

Luebbesmeyer, D.

2000-01-01

For the PANDA-Test-Facility (TEPSS configuration) post-test calculations and analyses have been performed for experiment -P2- (Early Start), -P3- (PCC start up), -P5- (Symmetric case, Two PCCs only) and -P7- (Severe Accident). Post test calculations have been performed with the system code RELAP5/Mod 3.2 using two different nodalization of the PANDA facility namely a basis nodalization and a much reduced one. The general trend of the calculations can be summarised: RELAP5/Mod3.2 calculated the general trends of the experiments sufficiently accurate; Using the reduced nodalization the results seem to be slightly more accurate than for the basic nodalization; On the other hand, calculations based on the reduced nodalization are not significantly faster than those with basic nodalization; The mass error is in the order of 200 to 900 kg. (author)

RELAP5/MOD2 assessment at Babcock and Wilcox

International Nuclear Information System (INIS)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Turk, C.

1986-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G Idaho, Inc. and the NRC assessing the RELAP5/MOD2 computer code by simulating selected separate effects tests. The purpose of this B and W Owners Group-sponsored assessment was to evaluate RELAP5/MOD2 for use in design calculations for the MIST and OTIS integral system tests and in predicting pressurized water reactor (PWR) transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (Cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve specific predictive capabilities of RELAP5/MOD2

Role of RELAP/SCDAPSIM in Nuclear Safety

Directory of Open Access Journals (Sweden)

C. M. Allison

2010-01-01

Full Text Available The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP. SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of some of the more important activities including the analysis of research reactors and Nuclear Power Plants, design and analysis of experiments, and training.

Simulation of water hammer experiments using RELAP5 code

International Nuclear Information System (INIS)

Kaliatka, A.; Vaisnoras, M.

2005-01-01

are presented. The influence of the control volume sizes in pipe components to the computational results is presented as well. The acquirement of knowledge will allow to develop RELAP5 code model for the analysis of accidents with the phenomenon of water hammer for the nuclear power plants. (author)

PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-04-01

rinci. Kata kunci: pemodelan, Generasi III+, RELAP5.   Westinghouse’s AP1000 reactor design is the first Generation III+ nuclear power reactor to receive final design approval from the U.S. Nuclear Regulatory Commission (NRC. Currently, the China’s utilities are starting construction several units of AP1000 on two selected sites for scheduled operation in 2013–2015. The AP1000, based on proven technology of Westinghouse-designed PWR with enhancement on the passive safety system, could be considered to be built in Indonesia referring to the requirements of government regulation No. 43/2006 regarding the Nuclear Reactor Licensing. To be accepted by the regulation agency, the design needs to be verified by independent Technical Support Organization (TSO, which can be done using RELAP5 computer code as accident analyses. Currently, NPP safety accident analysis is performed for PWR 1000 MWe of generation II or conventional type. Considering that nowadays references about the technology of AP1000 that includes passive safety technology has been available and assessed, a modeling activity used for future accident analyzes is introduced. Method for developing the model refers to IAEA guide consisting of plant data collection, engineering data and input deck development, and verification and validation of input data. The model developed should be considered preliminary but has been generally representing the AP1000 systems as the basic model. The model has been verified and validated by comparing thermalhidraulic parameter responses with design data in references with ± 13% deviation except for core pressure drop with 13% lower than design. As a basic model, the input deck is ready for further development by integrating safety system, protection system and control system model specified for AP1000 for purposes of safety simulation in detailed way. Keywords: Modeling, Generation III+ , RELAP5.

RELAP-7 Closure Correlations

Energy Technology Data Exchange (ETDEWEB)

Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, R. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, R. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hansel, J. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharpe, J. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Johns, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 code utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.

RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)

1995-09-01

Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

High Flux Isotope Reactor system RELAP5 input model

International Nuclear Information System (INIS)

Morris, D.G.; Wendel, M.W.

1993-01-01

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 reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

1991-01-01

A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

RELAP5 analyses of overcooling transients in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

1983-01-01

In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

Full Scope Modeling and Analysis on the Secondary Circuit of Chinese Large-Capacity Advanced PWR Based on RELAP5 Code

Directory of Open Access Journals (Sweden)

Dao-gang Lu

2015-01-01

Full Text Available Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developing process of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress. Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, training simulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPP will participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. In order to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 code has been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model of steady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalent method to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values are reasonable and acceptable.

RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

International Nuclear Information System (INIS)

Parisi, C.; D'Auria, F.

2007-01-01

The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)

RELAP5 based engineering simulator

International Nuclear Information System (INIS)

Charlton, T.R.; Laats, E.T.; Burtt, J.D.

1990-01-01

The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nuclear Plant Analyzer (NPA), have recently been improved significantly. This paper presents an overview of the INEL Engineering Simulation Center, and discusses the RELAP5/MOD3 and NPA/MOD1 codes, specifically how they are being used at the INEL Engineering Simulation Center. It provides an update on the modifications to these two codes and their application to the ATR engineering simulator project, as well as, a discussion on the reactor system representation, control system modeling, two phase flow and heat transfer modeling. It will also discuss how these two codes are providing desktop, stand-alone reactor simulation

Study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS using RELAP5 code

International Nuclear Information System (INIS)

Hoang Minh Giang; Ha Thi Anh Dao; Hoang Tan Hung; Bui Thi Hoa; Nguyen Thi Tu Oanh; Dinh Anh Tuan; Pham Tuan Nam

2017-01-01

The advanced VVER-1200/V491 reactor designed with passive safety systems to deal with design extension conditions is primarily selected as priority candidate for Ninh Thuan 1 nuclear power plant project. So that, in order to enhance competence of nuclear safety and toward participation on review Safety Analysis Report (SAR) of Ninh Thuan nuclear Power project the study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS is implemented. As requirement of the study, the input deck file of VVER-1200/V491 for RELAP5 and analysis report for some special case of LOCAs along with partly failure of ECCS are issued. (author)

Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

International Nuclear Information System (INIS)

Schneidesch, C.R.; Zhang Jinzhao

2004-01-01

A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

Extensions to SCDAP/RELAP5/MOD2 debris analysis models for the severe accident analysis of Savannah River Site (SRS) reactors preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.; Moore, R.L.

1989-06-01

Proposed extensions to the debris analysis model in the SCDAP/RELAP5 code to perform severe accident analyses of Savannah River Plant reactors are described. Designs are presented for the following areas of development: (a) calculating convective and radiative heat transfer at the surfaces of a debris region; (b) calculating heatup of a structure and supported debris that interfaces with several fluid control volumes; (c) modeling the addition of transported material to the surfaces of any structure represented by the debris analysis model; (d) calculating the two-dimensional heatup of an arbitrary number of structures in the reactor system; (e) modeling the effect of natural convection of liquefied material on heat transfer in a debris bed; and (f) modeling fission product release and aerosol generation in a debris bed. 11 refs., 12 figs., 7 tabs

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

2017-08-15

An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Directory of Open Access Journals (Sweden)

Takeshi Takeda

2017-08-01

Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

International Nuclear Information System (INIS)

Papini, Davide; Grgic, Davor; Cammi, Antonio; Ricotti, Marco E.

2011-01-01

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

Development of LabVIEW web-based simulator for RELAP5

Energy Technology Data Exchange (ETDEWEB)

Macedo, Luiz A.; Torres, Walmir M.; Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior Junior, Antonio; Umbehaun, Pedro E.; Conti, Thadeu N.; Mesquita, Roberto N. de; Masotti, Paulo H.F.; Angelo, Gabriel, E-mail: lamacedo@ipen.b, E-mail: wmtorres@ipen.b, E-mail: gdjian@ipen.b, E-mail: delvonei@ipen.b, E-mail: abelchior@ipen.b, E-mail: umbehaun@ipen.b, E-mail: tnconti@ipen.b, E-mail: rnavarro@ipen.b, E-mail: , E-mail: masotti@ipen.b, E-mail: gabriel.angelo@usp.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

This work presents the development of a LabVIEW web-based simulator using the output results of the best estimate nuclear system analysis code, RELAP5, for graphical user interfaces and web-casting. A numerical based model designed for natural circulation studies on the thermal hydraulic experimental facility called Natural Circulation Circuit, was developed with RELAP5 code. Specific output results from RELAP5 simulation are displayed in a user friendly graphical format. The temperatures are shown as a function of time in a XY graphic. Temperatures, levels and void fractions are displayed in color-coded scale which change in time on the graphical interface representing the circuit. An alarm is set for the case of onset boiling temperature occurrence at the heater outlet. This simulator allows an easy visual understanding of the thermal hydraulic circuit behavior. It can be shared, via Web, with researchers in any geographical location and, at the same time, it can be used in learning for distance educational purposes. In future work, this LabVIEW simulator will be coupled with RELAP5 code through dll's. Simultaneous graphical displaying and code calculations will be possible. Results are presented and discussed. (author)

Development of LabVIEW web-based simulator for RELAP5

International Nuclear Information System (INIS)

Macedo, Luiz A.; Torres, Walmir M.; Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior Junior, Antonio; Umbehaun, Pedro E.; Conti, Thadeu N.; Mesquita, Roberto N. de; Masotti, Paulo H.F.; Angelo, Gabriel

2011-01-01

This work presents the development of a LabVIEW web-based simulator using the output results of the best estimate nuclear system analysis code, RELAP5, for graphical user interfaces and web-casting. A numerical based model designed for natural circulation studies on the thermal hydraulic experimental facility called Natural Circulation Circuit, was developed with RELAP5 code. Specific output results from RELAP5 simulation are displayed in a user friendly graphical format. The temperatures are shown as a function of time in a XY graphic. Temperatures, levels and void fractions are displayed in color-coded scale which change in time on the graphical interface representing the circuit. An alarm is set for the case of onset boiling temperature occurrence at the heater outlet. This simulator allows an easy visual understanding of the thermal hydraulic circuit behavior. It can be shared, via Web, with researchers in any geographical location and, at the same time, it can be used in learning for distance educational purposes. In future work, this LabVIEW simulator will be coupled with RELAP5 code through dll's. Simultaneous graphical displaying and code calculations will be possible. Results are presented and discussed. (author)

SCDAP/RELAP5/MOD 3.1 code manual: User's guide and input manual. Volume 3

International Nuclear Information System (INIS)

Coryell, E.W.; Johnsen, E.C.; Allison, C.M.

1995-06-01

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 accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume provides guidelines to code users based upon lessons learned during the developmental assessment process. A description of problem control and the installation process is included. Appendix a contains the description of the input requirements

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

RELAP5-3D code validation for RBMK phenomena

International Nuclear Information System (INIS)

Fisher, J.E.

1999-01-01

The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena

User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

Energy Technology Data Exchange (ETDEWEB)

Forsmann, J. Hope [Idaho National Lab. (INL), Idaho Falls, ID (United States); Weaver, Walter L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

RELAP5 based engineering simulator

International Nuclear Information System (INIS)

Charlton, T.R.; Laats, E.T.; Burtt, J.D.

1990-01-01

The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor II (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nuclear Plant Analyzer (NPA), have recently been improved significantly. This paper will present an overview of the INEL Engineering Simulation Center, and discuss the RELAP5/MOD3 and NPA/MOD1 codes, specifically how they are being used at the INEL Engineering Simulation Center. It will provide an update on the modifications to these two codes and their application to the ATR engineering simulator project, as well as, a discussion on the reactor system representation, control system modeling, two phase flow and heat transfer modeling. It will also discuss how these two codes are providing desktop, stand-alone reactor simulation. 12 refs., 2 figs

Evaluation of void fraction measurements from DADINE experience using RELAP4/MOD5 code

International Nuclear Information System (INIS)

Borges, R.C.; Freitas, R.L.

1989-01-01

The DADINE experiment measures the axial evolution of the void fraction by neutronic diffusion in two-phase flow in the wet regions of a pressurized water reactor in accident conditions. Since the theoretical/experimental confrontation is important for code evaluation, this paper presents the simulation with the RELAP4/MOD5 Code of the void fractions results obtained in the DADINE Experiment, that showed some deviation probably associated with the existing models in Code, special attention in the way of stablishing the two-phase flow and the no characterization of the differents flow regimes related with the void fractions. (author) [pt

The calculation of coolant leak rate through the cracks using RELAP5 code

International Nuclear Information System (INIS)

Krungeleviciute, V.; Kaliatka, A.

2001-01-01

For reason to choose method of leak detection first of all it is necessary to perform evaluating thermal-hydraulic calculations. These calculations allow to determine flow rate of discharged coolant. For coolant leak rate calculations through possible cracks in Ignalina NPP pipes SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as computer program that predicts the leakage for cracked pipes in NPP. As this code calculates only water (at subcooled or saturated conditions) leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. For model validation comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows RELAP5 code model suitability for calculations of leak through through-wall cracks in pipes. (author)

Re-assessment of RELAP/SCDAPSIM/MOD3.X using historical integral experiments

Energy Technology Data Exchange (ETDEWEB)

Allison, Chris M.; Hohorst, Judith K. [Innovative Systems Software, Ammon, ID (United States)

2017-10-15

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed by Innovative Systems Software (ISS) as part of the international SCDAP Development and Training Program (STDP). The ISS developed RELAP/SCDAPSIM uses the publicly available SCDAP/RELAP5 models and correlations developed by the US Nuclear Regulatory Commission in combination with proprietary features developed by ISS and STDP members. Experimental versions contain improved models for LWRs including improved SCDAP models and correlations. This paper summarises and describes the new models that were incorporated into the experimental versions of RELAP/SCDAPSIM along with the extensive assessment and verification activities that are currently underway at ISS and various universities and institutes around the world along with examples of thee assessment results.

Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility

Energy Technology Data Exchange (ETDEWEB)

Kumari, Indu; Khanna, Ashok, E-mail: akhanna@iitk.ac.in

2017-04-01

Highlights: • Detail discussion of thermo physical properties of Lead Bismuth Eutectic incorporated in the code RELAP5/Mod4.0 included. • Benchmarking of LBE properties in RELAP5/Mod4.0 against literature. • NACIE facility for three different power levels (10.8, 21.7 and 32.5 kW) under natural circulation considered for benchmarking. • Preliminary validation of the LBE properties against experimental data. • NACIE facility for power level 22.5 kW considered for validation. - Abstract: The one-dimensional thermal hydraulic computer code RELAP5 was developed for thermal hydraulic study of light water reactor as well as for nuclear research reactors. The purpose of this work is to evaluate the code RELAP5/Mod4.0 for analysis of research reactors. This paper consists of three major sections. The first section presents detailed discussions on thermo-physical properties of Lead Bismuth Eutectic (LBE) incorporated in RELAP5/Mod4.0 code. In the second section, benchmarking of RELAP5/Mod4.0 has been done with the Natural Circulation Experimental (NACIE) facility in comparison with Barone’s simulations using RELAP5/Mod3.3. Three different power levels (10.8 kW, 21.7 kW and 32.5 kW) under natural circulation conditions are considered. Results obtained for LBE temperatures, temperature difference across heat section, pin surface temperatures, mass flow rates and heat transfer coefficients in heat section heat exchanger are in agreement with Barone’s simulation results within 7% of average relative error. Third section presents validation of RELAP5/Mod4.0 against the experimental data of NACIE facility performed by Tarantino et al. test number 21 at power of 22.5 kW comparing the profiles of temperatures, mass flow rate and velocity of LBE. Simulation and experimental results agree within 7% of average relative error.

RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads

International Nuclear Information System (INIS)

Stubbe, E.J.; VanHoenacker, L.; Otero, R.

1994-02-01

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 discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. The code results are compared to experimental load measurements performed at the Combustion Engineering Laboratory in Windsor (US). Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transients challenges the applicability of the following code models: two-phase choked discharge; interphase drag in conditions with large density gradients; heat transfer to metallic structures in fast changing conditions; two-phase flow at abrupt expansions. The code applicability to this kind of transients is investigated. Some sensitivity analyses to different code and model options are performed. Finally, the suitability of the code and some modeling guidelines are discussed

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

Development of loca calculation capability with relap5-3D in accordance with the evaluation model methodology

International Nuclear Information System (INIS)

Liang, T.K.S.; Huan-Jen, Hung; Chin-Jang, Chang; Lance, Wang

2001-01-01

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA (loss of coolant accident) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT (peak cladding temperature) evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by the Appendix K of 10 CFR 50 upon an advanced thermal-hydraulic platform can also enlarge significant margin between the highest calculated PCT and the safety limit of 2200 F. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in the Appendix K of 10 CFR 50. The associated models for LOCA consequent phenomenon analysis should follow the major concern of regulation and be expected to give more conservative results than those by the best-estimate methodology. They were required to predict the decay power level, the blowdown hydraulics, the blowdown heat transfer, the flooding rate, and the flooding heat transfer. All of the ten areas included in above classified simulations have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. In addition, to verify and assess the development of the Appendix K version of RELAP5-3D, nine separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with the Appendix K of 10 CFR 50 was demonstrated. We will apply another six sets of integral-effect experiments in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation. (authors)

Improvement of the RELAP5 subcooled boiling model for low pressure conditions

International Nuclear Information System (INIS)

Koncar, B.; Mavko, B.

2000-01-01

The RELAP5/MOD3.2.2 Gamma code was assessed against low pressure subcooled boiling experiments performed by Zeitoun and Shoukri [1] in a vertical annulus. The predictions of subcooled boiling bubbly flow showed that the present version of the RELAP5 code underestimates the void fraction growth along the tube. To improve the void fraction prediction at low pressure conditions a set of model changes is proposed, which includes modifications of bubbly-slug transition criterion, drift-flux model, interphase heat transfer coefficient and wall evaporation modeling. The improved experiment predictions with the modified RELAP5 code are presented and analysed. (author)

Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

International Nuclear Information System (INIS)

Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

1986-01-01

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

Application of RELAP/SCDAPSIM with integrated uncertainty options to research reactor systems thermal hydraulic analysis

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.; Perez, M.; Reventos, F.

2010-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. One such member developed option is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). This paper briefly summarizes the features of RELAP/SCDAPSIM/MOD4.0 and the integrated uncertainty analysis package, and then presents an example of how the integrated uncertainty package can be setup and used for a simple pipe flow problem. (author)

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P. [Nuclear Research Inst. Rez (Switzerland)

1995-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P [Nuclear Research Inst. Rez (Switzerland)

1996-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

International Nuclear Information System (INIS)

Kral, P.

1995-01-01

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

Development of Multi-Dimensional RELAP5 with Conservative Momentum Flux

Energy Technology Data Exchange (ETDEWEB)

Jang, Hyung Wook; Lee, Sang Yong [KINGS, Ulsan (Korea, Republic of)

2016-10-15

The non-conservative form of the momentum equations are used in many codes. It tells us that using the non-conservative form in the non-porous or open body problem may not be good. In this paper, two aspects concerning the multi-dimensional codes will be discussed. Once the validity of the modified code is confirmed, it is applied to the analysis of the large break LOCA for APR-1400. One of them is the properness of the type of the momentum equations. The other discussion will be the implementation of the conservative momentum flux term in RELAP5. From the present study and former, it is shown that the RELAP5 Multi-D with conservative convective terms is applicable to LOCA analysis. And the implementation of the conservative convective terms in RELAP5 seems to be successful. Further efforts have to be made on making it more robust.

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

1997-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N [St. Petersburg State Technical Univ. (Russian Federation); Banati, J [Lappeenranta Univ. of Technology (Finland)

1998-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design 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 RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

Summary of severe accident assessment for Atucha 2 Nuclear Power Plant using RELAP5/SCDAPSIM Mod3.6

International Nuclear Information System (INIS)

Bonelli, Analia; Mazzantini, Oscar; Siefken, Larry; Allison, Chris

2014-01-01

A severe accident assessment was performed for the Atucha 2 Nuclear Power Plant in Argentina. Atucha 2 is a PHWR, cooled and moderated by heavy water, presently in commissioning process. Its 451 fuel assemblies are 6.03m high and each composed of 37 Zircaloy clad fuel rods. Each assembly is placed inside an individual Zircaloy coolant channel. Heavy water coolant flows inside the channels which are all immersed inside the moderator tank. The RPV lower plenum is occupied by a massive steel structure called 'filling body' that was designed to minimize heavy water inventory. Due to some unique design characteristics, severe accident progression in Atucha 2 is expected to be somewhat different from that predicted for regular PWRs. Therefore, a very detailed assessment was performed, focused on the different accident stages and expected phenomena by the use of different input models and nodalizations. When possible, linking to available experimental data was performed. RELAP/SCDAPSIM Mod 3.6 was the computer code selected to perform this task. The modeling of Atucha 2's unique characteristics required several extensions to the code. For the severe accident assessment of Atucha 2, three different input models were developed that were key instruments for the debugging and evaluation process. A Single Channel Model was used to evaluate the first stages of core heatup (including the boiloff of the channels and moderator tank), an RPV standalone model was used to assess the interaction between components in the complete core and for the evaluation of late in-core melting and relocation. Then, a Lower Plenum standalone model was developed to assess the behavior of the melted and slumped core material on top of the filling body and to analyze ex-vessel cooling as a possible severe accident management action. For each of the cases, highlights of key results are shown and general conclusions are drawn. In the case of a severe accident with significant meltdown of

PHISICS multi-group transport neutronic capabilities for RELAP5

Energy Technology Data Exchange (ETDEWEB)

Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

2012-07-01

PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Supplement 1, RELAP4/MOD5, Update 2

International Nuclear Information System (INIS)

Bruch, C.G.

1976-08-01

RELAP4/MOD5, Update 1 was released to the Nuclear Regulatory Commission in January 1976. Six months of extensive use of Update 1 has led to the identification and correction of a number of errors in the code, as well as some logic changes, additional Evaluation Model (EM) checks, and one model revision. These changes have culminated in the release of an improved code identified as RELAP4/MOD5, Update 2. The RELAP4/MOD5 interim User's Manual (Interim Report SRD-113-76) reflected the Update 1 version of the code. The purpose of the supplement presented is to update the Interim User's Manual for use with RELAP4/MOD5, Update 2. Major differences between Updates 1 and 2 and the checkout of Update 2 are discussed. The final version of the User's Manual will be written in accordance with Update 2 and will be published as ANCR-NUREG 1335 during September 1976. Annotation of the existing three volumes of the User's Manual to cross-reference this Supplement is recommended

Quality control of the packet of RELAP5/MOD2 code

International Nuclear Information System (INIS)

Pomier Baez, L.E.

1993-01-01

The methodology that should be used to perform the quality control of entrance data set of RELAP5 calculation code is expounded in this work with this control method an extreme reliability is quarantined in the calculation model established to perform the safety thermohydraulic analysis with the help of RELAP5. This makes possible the complex simulation studies of a nuclear power plant with the quality required

International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

International Nuclear Information System (INIS)

Schultz, R.R.

1993-12-01

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 discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

RELAP5/MOD3 analysis of a heated channel in downflow

International Nuclear Information System (INIS)

Dimenna, R.A.; Qureshi, Z.H.; Boman, A.L.

1993-01-01

The onset of flow instability (OFI) is a significant phenomenon affecting the determination of a safe operating power limit in the Savannah River Site production reactors. Tests performed at Columbia University for a single tube with uniform axial and azimuthal heating have been analyzed with RELAP5/NPR, Version 0, a version of RELAP5/MOD3. The tests include water flow rates from 3.2 x l0 -4 - 2.l x 10 -3 m 3 /s (5 - 33 gpm), Reynolds numbers from 30,000 - 400,000, and surface heat fluxes from 0 - 3.2 x l0 6 w/m 2 (0 - 1,000,000 Btu/hr- ft 2 ). Pressure drop versus flow rate curves were mapped for both fixed pressure boundary conditions and fixed flow boundary conditions. RELAP5/MOD3 results showed fair agreement with data for both types of boundary conditions, and good internal consistency between calculations using the two different types of boundary conditions. Under single-phase unheated conditions, the code overpredicted the pressure drop by 22 - 34%. Under single-phase heated conditions, the overprediction increased to as much as 55%. For those tests where two-phase conditions were observed at the channel exit, RELAP5 predicted lower flows than seen in the tests before voiding occurred

Extensions to the SCDAP/RELAP5 code for the modeling of debris oxidation and materials interactions preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.; Davis, K.L.

1993-02-01

Preliminary designs are proposed for extending the SCDAP/RELAP5 code so that it models (a) the oxidation of slumping fuel rod material and cohesive and porous debris and (b) the interaction of PWR control rod materials with the other materials in a reactor core. These extensions have the purpose of improving the code's calculation of the damage progression and hydrogen production that takes place during the early phase of a severe accident

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

International Nuclear Information System (INIS)

Bovalini, R.; D'Auria, F.; Galassi, G.M.; Mazzini, M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool of ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions

Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

International Nuclear Information System (INIS)

Keco, M.; Debrecin, N.; Grgic, D.

2005-01-01

Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2017-01-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1980-06-01

A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author) [pt

Assessment of RELAP5/MOD2 and RELAP5/MOD1-EUR codes on the basis of LOBI-MOD2 test results

International Nuclear Information System (INIS)

D'Auria, F.; Mazzini, M.; Oriolo, F.; Galassi, G.M.

1989-10-01

The present report deals with an overview of the application of RELAP5/MOD2 and RELAP5/MOD1-EUR codes to tests performed in the LOBI/MOD2 facility. The work has been carried out in the frame of a contract between Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN) of Pisa University and CEC. The Universities of Roma, Pisa, Bologna and Palermo and the Polytechnic of Torino performed the post-test analysis of the LOBI experiment under the supervision of DCMN. In the report the main outcomes from the analysis of the LOBI experiments are given with the attempt to identify deficiencies in the modelling capabilities of the used codes

RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

1984-01-01

In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

SCDAP/RELAP5 applications to RCS natural circulation

International Nuclear Information System (INIS)

Bayless, P.D.

1988-01-01

The effects of natural circulation flows in the reactor coolant system during a TMLB' sequence were investigated. Both in-vessel circulation and hot leg countercurrent flow were modeled in the Surry nuclear power plant using the SCDAP/RELAP5 computer code. The transient was analyzed until after fuel rod relocation had begun. The delays in the onset of relocation resulting from the natural circulation flows were not significant compared to SCDAP/RELAP5 calculations without natural circulation modeled, but were large compared to the analyses presented in NUREG-1150. The most significant aspect of the natural circulations flows was the heating of ex-vessel structures. Surge line failure is likely to occur before the vessel is breached by the molten core, while steam generator tube failure is not expected

RL5SORT/RL5PLOT. A graphite package for the JRC-Ispra IBM version of RELAP5/MOD1

International Nuclear Information System (INIS)

Kolar, W.; Brewka, W.

1984-01-01

The present report describes the programs RL5SORT and RL5PLOT, their implementation and ''how to use''. Both programs base on the IBM version of RELAP5 as developed at JRC-ISPRA. RL5SORT creates from the output file (restart plot file) of a RELAP5 calculation data files, which serve as input data base for the program RL5PLOT. RL5PLOT retrieves the previous stored data records (minor edit quantities of RELAP5), allows arithmetic operations with the retrieved data and enables a print or graphic output on the terminal screen of a TEKTRONIX graphic terminal. A set of commands, incorporated in the program RL5PLOT, facilitates the user's work. Program RL5SORT has been developed as a batch program, while RL5PLOT has been conceived for interactive working mode

Lesson learned from the application to LOBI tests of CATHARE and RELAP5 codes

International Nuclear Information System (INIS)

Ambrosini, W.; D'Auria, F.; Galassi, G.M.

1992-01-01

The Dipt. di Costruzioni Meccaniche e Nucleari has participated to the LOBI project since its very beginning, contributing to almost all the international activities in this field, such as task group meetings, International Standards Problems, Seminars, etc. System codes like RELAP4/MOD6, RELAP5/MOD1, RELAP5/MOD1-EUR, RELAP5/MOD2, CATHARE 1 and CATHARE 2 were applied to the design and post test evaluation of a wide series of both LOBI/MOD1 and LOBI/MOD2 experiments, including Large Break LOCAs, Small and Intermediate Break LOCAs, long lasting transients and characterization tests. The LOBI data base demonstrated its usefulness in assessing capabilities and limitations of these codes and in qualifying a code use strategy. (author)

Advanced Presentation of BETHSY 6.2TC Test Results Calculated by RELAP5 and TRACE

Directory of Open Access Journals (Sweden)

Andrej Prošek

2012-01-01

Full Text Available Today most software applications come with a graphical user interface, including U.S. Nuclear Regulatory Commission TRAC/RELAP Advanced Computational Engine (TRACE best-estimate reactor system code. The graphical user interface is called Symbolic Nuclear Analysis Package (SNAP. The purpose of the present study was to assess the TRACE computer code and to assess the SNAP capabilities for input deck preparation and advanced presentation of the results. BETHSY 6.2 TC test was selected, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The RELAP5 legacy input deck was converted to TRACE input deck using SNAP. The RELAP5 and TRACE comparison to experimental data showed that TRACE results are as good as or better than the RELAP5 calculated results. The developed animation masks were of great help in comparison of results and investigating the calculated physical phenomena and processes.

Development of a Wrapper Object, TRelap, for RELAP5 Code for Use in Object Oriented Programs

International Nuclear Information System (INIS)

Lee, Young Jin

2008-01-01

TRelap object class has been developed to enable object oriented programming techniques to be used where functionality of the RELAP5 thermal hydraulic system analysis code is needed. The TRelap is an object front for Dynamic Link Library (DLL) manifestation of the Relap5 code, Relap5.dll. In making the Relap5.dll, the top most structure of the RELAP5 was altered to enable the external calling procedures to control and the access the memory. The alteration was performed in such a way to allow the entire 'fa' and the f tb' memory spaces to be accessible to the calling procedure. Thus, any variable contained within the 'fa' array such as the parameters for the components, volumes, junctions, and heat structures can be accessed by the external calling procedure through TRelap. Various methods and properties to control the RELAP5 calculation and to access and manipulate the variables are built into the TRelap to enable easy manipulation. As a verification effort, a simple program was written to demonstrate the capability of the TRelap

NPP Krsko containment environmental conditions during postulated accident

International Nuclear Information System (INIS)

Kozaric, M.; Cavlina, N.; Spalj, S.

1989-01-01

This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

Summary description of the RELAP5 Koeberg-1 simulation model

International Nuclear Information System (INIS)

D'Arcy, A.J.

1990-11-01

The main features of the RELAP5 code and the model are summarized. The model has been quality-assured in accordance with a QA programme used in the Reactor Theory Group of the Atomic Energy Corporation of SA Ltd. The RELAP5 code is based on a non-homogeneous, non-equilibrium model for the two-phase system that is solved by a fast, partially implicit numerical scheme. The objective of the development effort from the outset has been to produce a code that includes important first-order effects necessary for accurate prediction of system transients, but is sufficiently simple and cost-effective so that parametric or sensitivity studies are possible. The code includes many generic component models from which general systems can be simulated. Special process models are included for effects such as form losses, flow at an abrupt area change, branching, choked flow, boron tracking and a non-condensible gas. The RELAP5 modelling and computational aspects covered are: hydrodynamic models, constitutive package, special process models, and user conveniences. 25 tabs., 8 figs., 17 refs

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

International Nuclear Information System (INIS)

Andrs, David; Berry, Ray; Gaston, Derek; Martineau, Richard; Peterson, John; Zhang, Hongbin; Zhao, Haihua; Zou, Ling

2012-01-01

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

Code development and analysis program. RELAP4/MOD7 (Version 2): user's manual

International Nuclear Information System (INIS)

1978-08-01

This manual describes RELAP4/MOD7 (Version 2), which is the latest version of the RELAP4 LPWR blowdown code. Version 2 is a precursor to the final version of RELAP4/MOD7, which will address LPWR LOCA analysis in integral fashion (i.e., blowdown, refill, and reflood in continuous fashion). This manual describes the new code models and provides application information required to utilize the code. It must be used in conjunction with the RELAP4/MOD5 User's Manual (ANCR-NUREG-1335, dated September 1976), and the RELAP4/MOD6 User's Manual

Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

International Nuclear Information System (INIS)

Preece, R.J.; Putney, J.M.

1993-07-01

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

Safety assessment of the Indonesian multipurpose reactor RSG-GAS against ATWS and hypothetical accidents

International Nuclear Information System (INIS)

Hastowo, H.; Nabbi, R.; Prayoto; Ismuntoyo, R.P.H.

2004-01-01

Investigation on ATWS and hypothetical accidents for the Indonesian Multipurpose Reactor RSG-GAS have been undertaken by computer simulation technique. Two computer codes, namely RELAP5 and PARET-ANL, were used as the main tools. The RELAP5 was utilized to perform system analysis while the PARET-ANL code was used to perform the reactor core analysis in more detail. Two different models have been applied as a basis of the simulation: Typical Working Core model (IWC-model) consisting of four regions with different radial power factors; and the hot-channel model consisting of two regions with different radial power factors. Both RELAP5 ad PARET-ANL results showed that in the occurrence of ATWS, failure on fuel element or fuel plate was limited to the region with the most highest power factor. The results also indicated that no high pressure development occurs in that region, so that mechanical damage on the fuel element or other core components due to pressure shock did not happen.(author)

Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

1992-01-01

The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons

Vectorization and improvement of nuclear codes (MEUDAS4, FORCE, STREAM V2.6, HEATING7-VP, SCDAP/RELAP5/MOD2.5, NBI3DGFN)

International Nuclear Information System (INIS)

Nemoto, Toshiyuki; Suzuki, Koichiro; Isobe, Nobuo; Machida, Masahiko; Osanai, Seiji; Yokokawa, Mitsuo

1992-09-01

Eight nuclear codes have been vectorized and modified to improve their performance. These codes are magnetic fluid equilibrium code MEUDAS4 (CR and FFT versions), the magnetic field analysis code FORCE, the three-dimensional heat fluid analysis code STREAM V2.6, the three-dimensional heat analysis code HEATING 7-VP, the severe accident transient analysis code SCDAP/RELAP 5/MOD 2.5 for light water reactors, the ion beam orbital analysis code NBI3DGFN, and a free electron laser analysis code. The speedup ratios of the vectorized versions to the original ones in scalar mode are 2.3-4.9, 1.9-5.4, 2.6-6.2, and 1.9 for the MEUDAS4, STREAM, FORCE, and free electron laser analysis code, respectively. The definition method of the computational regions in the HEATING7-VP is improved. The SCDAP/RELAP5/MOD2.5 is modified to use extended memory regions of the computer. In this report, outlines of the codes, techniques used in the vectorization and reorganization of the codes, verification of computed results, and improvement on the performance are presented. (author)

A generic semi-implicit coupling methodology for use in RELAP5-3Dcopyright

International Nuclear Information System (INIS)

Aumiller, D.L.; Tomlinson, E.T.; Weaver, W.L.

2000-01-01

A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3Dcopyright computer program. This methodology allows RELAP5-3Dcopyright to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered. The methodology was demonstrated using a test case in which the test geometry was divided into two parts each of which was solved as a RELAP5-3Dcopyright simulation. This test problem exercised all of the semi-implicit coupling features which were installed in RELAP5-3D0. The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

PRELIMINARY STUDY ON RELAP5 SIMULATION OF DVI LINE BREAK ACCIDENT IN THE ATLAS FACILITY USING BEST ESTIMATE PLUS UNCERTAINTY METHOD

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2017-03-01

STUDI AWAL SIMULASI KECELAKAAN PUTUSNYA JALUR DVI PADA FASILITAS ATLAS MENGGUNAKAN RELAP5 DENGAN METODE ESTIMASI TERBAIK DAN KETIDAKPASTIAN. Metode Best estimate plus uncertainty (BEPU adalah metode analisis keselamatan deterministik yang bertujuan untuk melakukan evaluasi keterbatasan program perhitungan dalam mensimulasikan sifat-sifat fisis instalasi secara realistik dengan mengkuantifikasi rentang ketidakpastian dari hasil perhitungan. Metode tersebut telah diterima secara luas dalam perijinan PLTN oleh badan pengatur dunia seperti di Amerika (USNRC, di Argentina, dan Kanada. Evaluasi ketidakpastian dalam metode BEPU dilakukan dengan beberapa metode yang berbeda seperti GRS, IRSN, ENUSA, AEAT, dan UNIPI. Atas dasar kompleksitas metode-metode yang lain, tujuan makalah ini adalah untuk menggambarkan aspek penting dari proses BEPU dengan metode GRS dengan melakukan simulasi putusnya jalur DVI sebesar 50% luasan pada fasilitas ATLAS karena analisis keselamatan yang dilakukan selama ini baru berupa perkiraan terbaik secara deterministik. Sebagai perbandingan dari simulasi perkiraan terbaik yang dilakukan dengan RELAP5/SCDAP/Mod3.4 digunakan data-data eksperimen yang telah terdokumentasi. Setelah dilakukan 100 simulasi, rentang ketidakpastian dari transien temperatur puncak kelongsong pemanas dan tekanan primer hanya mendekati data eksperimen pada 250 detik di periode awal. Oleh karena itu keakuratan dari simulasi perkiraan terbaik secara keseluruhan memiliki peranan penting pada hasil akhir dari analisis ketidakpastian karena perambatan perbedaan dengan data eksperimen akan terus terjadi selama simulasi. Setelah itu, pemilihan parameter yang penting untuk dikembangkan secara random harus dilakukan secara cermat dengan mempelajari fenomena-fenomena penting yang terkait dengan kejadian yang dianalisis dan model instalasinya. Kata kunci: perkiraan terbaik dan ketidakpastian, putusnya jalur DVI, fasilitas ATLAS, RELAP5, simulasi

Steady state HTTR model in the RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Scari, Maria E.; Reis, Patrícia A. L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A. F., E-mail: melizabethscari@yahoo.com, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPQ (Brazil)

2017-07-01

The HTTR (High Temperature Engineering Test Reactor) is a high temperature gas-cooled reactor (HTGR). It is a helium-cooled and graphite-moderated with a thermal power of 30 MW, developed and operated by the Japan Atomic Energy Research Institute (JAERI). In a previous work, a thermal model for the HTTR has been developed using the RELAP5-3D code. In the present work, additional thermal analyses have been presented. Studies about the core neutronic simulation were also initiated and are described here. The code WIMSD-5B was used to generate the cross sections and the group constants necessary to the NESTLE neutron kinetic code incorporated in the RELAP5-3D. (author)

Gest-sip1 experiments and post-test calculations with the relap5 code

International Nuclear Information System (INIS)

Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S.; Bianchi, F.; Meloni, P.

2001-01-01

The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

Simulation of a loss of primary coolant accident due to a large break in Angra 2 Nuclear Power Plant with RELAP5/MOD3.2.2G code

International Nuclear Information System (INIS)

Sabunddjian, Gaiane; Andrade, Delvonei Alves de

2003-01-01

This work presents the simulation, with RELAP5/MOD.3.2.2G code, of the postulate accident with loss of coolant in the primary circuit for large break (LBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 FSAR. The accident consists basically of the total break of the cold leg (Loop 20) of Angra 2 Plant. The rupture area considered is 4418 cm 2 , which represents 100% of the primary circuit pipe flow area. The Emergency Core Cooling System (ECCS) efficiency is also verified for this accident. In this simulation, failure and repair criteria are adopted for the ECCS components, in order to verify the system operation, in carrying out its function as expected by the project to preserve the integrity of the reactor core and to guarantee its cooling. LBLOCA accidents are characterized by a fast blowdown in the primary circuit to values that the low pressure injection system is activated and then, followed by the water injection by the accumulators. The thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the core liquid level, until the ECCS is capable to reflood it. It is important to point out that the results do not represent an independent calculation for the licensing process, but a calculation to give support to the qualification process of Angra 2 transient basic nodalization (author)

RELAP5/MOD2 development

International Nuclear Information System (INIS)

Miller, C.S.

1986-01-01

Status of the RELAP5/MOD2 computer code is discussed. While the code is undergoing international assessment, emphasis is on user support and code maintenance with modifications allowed only for error correction and user convenience improvements. User support discussed is the response to user inquiries, maintenance of manuals and the implementation of a PC based newletter service. The major 1986 user convenience improvement is the self-initialization option. The method is discussed and examples for PWR ''U tube'' and ''once through'' plants are illustrated. Future plans are also outlined

RELAP5/MOD2 development

International Nuclear Information System (INIS)

Miller, C.S.

1987-01-01

Status of the RELAP5/MOD2 computer code is discussed. While the code is undergoing international assessment, emphasis is on user support and code maintenance with modifications allowed only for error correction and user convenience improvements. User support discussed is the response to user inquiries, maintenance of manuals and the implementation of a PC based newsletter service. The major 1986 user convenience improvement is the self-initialization option. The method is discussed and examples for PWR U tube and once through plants are illustrated. Future plans are also outlined

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

Assessment of RELAP5/MOD3 with condensation experiment for pure steam condensation in a vercal tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Jae; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

The film condensation models in RELAP5/MOD3.1 and RELAP5/MOD3.2 are assessed with the data of experiment performed in the scaled down condensation experimental facility with a single vertical tube of inner diameter of 46 mm in the range of pressure 0.1 {approx} 7.5 MPa for the PSCS(Passive Secondary Condenser System). Both MOD3.1 and MOD3.2 don`t shows any reliable predictions of the experimental data. The RELAP5/MOD3.1 overpredicts the heat transfer coefficients of experiment, whereas the RELAP5/MOD3.2 underpredicts those data. It is recommended that the film condensation model in RELAP5/MOD3.2 should be modified to have a larger heat transfer coefficient than those of the present model to give the reliable predictions. 7 refs., 6 figs., 1 tab. (Author)

Assessment of RELAP5/MOD3 with condensation experiment for pure steam condensation in a vercal tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Jae; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The film condensation models in RELAP5/MOD3.1 and RELAP5/MOD3.2 are assessed with the data of experiment performed in the scaled down condensation experimental facility with a single vertical tube of inner diameter of 46 mm in the range of pressure 0.1 {approx} 7.5 MPa for the PSCS(Passive Secondary Condenser System). Both MOD3.1 and MOD3.2 don`t shows any reliable predictions of the experimental data. The RELAP5/MOD3.1 overpredicts the heat transfer coefficients of experiment, whereas the RELAP5/MOD3.2 underpredicts those data. It is recommended that the film condensation model in RELAP5/MOD3.2 should be modified to have a larger heat transfer coefficient than those of the present model to give the reliable predictions. 7 refs., 6 figs., 1 tab. (Author)

Comparison of MAAP4.03 with RELAP/SCDAPSIM/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Kohriyama, T.; Ohtani, M. [Technical Support Project, Institute of Nuclear Technology, Institute of Nuclear Safety System, Incorporated, Mihama, Fukui (Japan); Ezzidi, A.; Morota, H. [Computer Software Development Co., Ltd., Tokyo (Japan)

2000-11-01

The MAAP4.03 code has been widely used for analyses of severe accident phenomena. It is, however, a system level code applying simpler models and could show phenomenological uncertainties. In order to support MAAP4.03 by a more detailed mechanistic code such as RELAP/SCDAPSIM/MOD3.2, code-to-code comparisons are performed. For a typical 4 loop PWR, analyses of two severe accident sequences were executed. Both codes predicted similar tendencies until the beginning of core melt. MAAP4.03 showed earlier slumping of molten core to a lower head of a reactor pressure vessel than RELAP/SCDAPSIM/MOD3.2. MAAP4.03 considers only axial flows of molten core and crusts suddenly breach by Creep Rupture. RELAP/SCDAPSIM/MOD3.2 treats axial and radial spreads by repeated cycles of melting, flowing and freezing. Bottom crusts can be supported by intact fuel rods. By these more realistic RELAP/SCDAPSIM/MOD3.2 models, MAAP4.03 could be supported maintaining conservatism. (author)

Development of a VBA macro-based spreadsheet application for RELAP5 data post-processing

International Nuclear Information System (INIS)

Belchior Junior, Antonio; Andrade, Delvonei A.; Sabundjian, Gaiane; Macedo, Luiz A.; Angelo, Gabriel; Torres, Walmir M.; Umbehaun, Pedro E.; Conti, Thadeu N.; Bruel, Renata N.

2011-01-01

During the use of thermal-hydraulic codes such as RELAP5, large amount of data has to be managed in order to prepare its input data and also to analyze the produced results. This work presents a helpful tool developed to make it easier to handle the RELAP5 output data file. The XTRIP application is an electronic spreadsheet that contains some programmed macros that should be used for post-processing the RELAP5 output file. It can directly read the RELAP5 restart-plot binary output file and, through a user-friendly interface, transient results can be chosen and exported directly into an electronic worksheet. The XTRIP program can also do some data unit conversion as well as export these data to other programs such as Wingraf, Grapher and COBRA, etc. The main features of the developed Excel Visual Basic for Application macro as well as an example of use are presented and discussed. (author)

Calculation of pre and post-test of the third. proposed standard problem exercise, for the PMK-NVH-IAEA experiment using the RELAP4/MOD5 and RELAP5/MOD1

International Nuclear Information System (INIS)

Neves Conti, T. das; Sabundjian, G.; Oliveira Neto, J.M. de

1992-01-01

The results of RELAP4/MOD5 and RELAP5/MOD1 modeling tests against the steam generator tube rupture experiments performed at PMK-NVH Experimental Loop Facility (IAEA-Standard Problem Exercise-3) are presented in the report. The pre and post-test results, when compared against the experimental data were satisfactorily good, except a discrepancy in the steam-generator relief valve opening time. (author)

RELAP5/MOD3.3 assessment against MSIV closure events in Krsko NPP

International Nuclear Information System (INIS)

Parzer, I.

2002-01-01

The paper presents RELAP5/MOD3.3 analysis of two abnormal events occurred in Krsko NPP originating from sudden closure of Main Steam Isolation Valve (MSIV). Both events occurred before the SG replacement in 2000, the first one in September 1995 and the second one in January 1997. Valuable plant data were obtained from real plant transients and the RELAP5 code assessment was performed. Recently the last frozen version RELAP5/MOD3.3 has been released, before merging with another best-estimate thermalhydraulic system code TRAC into an integrated code. It is thus of utmost importance to assess models built in RELAP5 code against real plant transients before the code merger. A full twoloop plant model, developed at Jozef Stefan Institute (JSI), has been used for the analyses. The model includes old Westinghouse D4 type steam generators (SGs) with assumed 18% Utubes plugged in both steam generators. In the first case a malfunction in the MSIV in SG-1 caused inadvertent valve closure, while in the second case the valve stem has been broken in the SG-2, which also caused sudden valve closure.(author)

Peer review of RELAP5/MOD3 documentation

International Nuclear Information System (INIS)

Craddick, W.G.

1993-01-01

A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Volume 3, Developmental Assessment problems, and Volume 4, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Volume 6, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax's Equivalence Theorem

RELAP-7 Numerical Stabilization: Entropy Viscosity Method

Energy Technology Data Exchange (ETDEWEB)

R. A. Berry; M. O. Delchini; J. Ragusa

2014-06-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL's modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5's capability and extends the analysis capability for all reactor system simulation scenarios. RELAP-7 utilizes a single phase and a novel seven-equation two-phase flow models as described in the RELAP-7 Theory Manual (INL/EXT-14-31366). The basic equation systems are hyperbolic, which generally require some type of stabilization (or artificial viscosity) to capture nonlinear discontinuities and to suppress advection-caused oscillations. This report documents one of the available options for this stabilization in RELAP-7 -- a new and novel approach known as the entropy viscosity method. Because the code is an ongoing development effort in which the physical sub models, numerics, and coding are evolving, so too must the specific details of the entropy viscosity stabilization method. Here the fundamentals of the method in their current state are presented.

Analysis of the reflood experiment by RELAP5/MOD2 code

International Nuclear Information System (INIS)

Prosek, A.; Stritar, A.

1990-01-01

The analysis of the reflood experiment on the test rig Achilles has been performed. The analysis has been done by the RELAP5/MOD2 code after the results of the experiment had been released. The experiment has been analyze in several other laboratories around the world. Our results are comparable to other analyses and are in the range of RELAP5/MOD2 capabilities. Two analyses have been done: the core only and the complete system. Computed clad temperatures in the first case are higher than measured, in the second case they are somewhat lower. (author)

Integrated uncertainty analysis using RELAP/SCDAPSIM/MOD4.0

International Nuclear Information System (INIS)

Perez, M.; Reventos, F.; Wagner, R.; Allison, C.

2009-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalunya (UPC) and Innovative Systems Software (ISS). The integrated uncertainty analysis approach used in the package uses the following steps: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. The first four steps are performed by the user prior to the RELAP/SCDAPSIM/MOD4.0 analysis. The remaining steps are included with the MOD4.0 integrated uncertainty analysis (IUA) package. This paper briefly describes the integrated uncertainty analysis package including (a) the features of the package, (b) the implementation of the package into RELAP/SCDAPSIM/MOD4.0, and

Improvements to the RELAP/SCDAPSIM of Laguna Verde model for the analysis of transients and accidents; Mejoras al modelo de Laguna Verde de RELAP/SCDAPSIM para el analisis de transitorios y accidentes

Energy Technology Data Exchange (ETDEWEB)

Amador G, R.; Castillo D, R.; Ortiz V, J.; Araiza M, E.; Martinez C, E., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

This work presents the improvements to the integral model of the nuclear power plant of Laguna Verde for the RELAP/SCDAPSIM code, for the simulation of transients and severe accidents. The model includes a new detailed geometry of the steam lines, as well as improvements in the performance of the emergency systems. A primary containment model has also been created, which will be used to analyze the effect of safety valve and relief valve discharges to the wet well suppression pool and the effect of the rupture of a recirculation loop on the dry well. The simulations performed with the new model show that the changes made improve the prediction of the phenomenology involved during transients and accidents. (Author)

RELAP5/MOD2 Overview and Developmental. Assessment Results from TMl-1 Plant Transient Analysis

International Nuclear Information System (INIS)

Lin, J. C.; Tsai, C. C.; Ransom, V. H.; Johnsen, G. W.

2013-01-01

RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. The objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly sub-cooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2. (author)

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

RELAP5/MOD2 code assessment for the Semiscale Mod-2C Test S-LH-1

International Nuclear Information System (INIS)

Fineman, C.P.

1986-01-01

RELAP5/MOD2, Cycle 36.02, was assessed using data from Semiscale Mod-2C experiment S-LH-1. The major phenomena that occurred during the experiment were calculated by RELAP5/MOD2, although the duration and the magnitude of their effect on the transient were not always well calculated. Areas defined where further work was needed to improve the RELAP5 calculation include: (1) the system energy balance, (2) core interfacial drag, and 3) the heat transfer logic rod dryout criterion

Presentation of RELAP5 results on the personal computer

International Nuclear Information System (INIS)

Salamun, I.; Stritar, A.

1991-01-01

DrALF is a program for graphical presentation of RELAP5 results. Results may be displayed in two different forms, as graphs with different zoom capabilities and as drawings or nodalizations with different variables displayed on a background picture. (author)

The gradual development steps of the external coupled RELAP5 - DYN3D code

International Nuclear Information System (INIS)

Strmensky, C.

2001-01-01

This paper describes the on-going and finished parts of project: 'The external coupled RELAP5-DYN3D code'. The development progress was divided into four steps. In present time, second and third steps are performed and four step is started. The two parameters of coolant was selected and are exchanged between codes RELAP5 and DYN3D. (authors)

Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

International Nuclear Information System (INIS)

Atanasova, B.; Grudev, P.

2011-01-01

The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

Investigation of 3D spatial effect on point kinetics estimation of the thermal hydraulics code RELAP for the analysis of MSLB accident of KK-NP

International Nuclear Information System (INIS)

Bera, S.; Pradhan, S.K.; Dubey, S.K.; Gupta, S.K.

2011-01-01

In general safety analyses of design basis accident of NPPs are being carried out using system thermal hydraulics code like RELAP. In RELAP, power is calculated based on point kinetics approximation, which virtually ignores the space and energy dependence of neutron flux. To include the space and energy dependence of neutron flux, three-dimensional neutronics code TRIHEXFA has been externally coupled with RELAP through interface program, TRIHEXFA-RELAP Interface Program (TRIP). Calculation methodology of TRIP program is based on adiabatic approximation. In the adiabatic approximation the neutron flux is being factored into spatial and amplitude part. Spatial part of flux is slowly varying with time whereas amplitude part is strongly varying function. The RELAP controls the transient time steps. Transient time is divided into several major and minor time steps. Minor time step is the sub-step of major time step. Thermal hydraulics and neutronics data are exchanged at each major time step. Spatial part of neutron flux has been updated at each major time step using TRIHEXFA code. But amplitude part of the neutron flux is calculated at each minor time step using RELAP code. Convergence of results of the coupled code, TRIP has been checked through coupling time step descritization study. This study determines the minimum coupling time step. Transient concerning VVER-1000 Main Steam Line Break, MSLB has been considered to investigate the space-time effect on point kinetics. MSLB occurs as a consequence of the rupture of one steam line upstream of main steam line isolation valves. Reference design and data from Kudankulam Nuclear Power Plant (KK-NPP) are used for the analysis. From this investigation it is found that TRIP significantly overestimates the maximum reactor power against uncoupled RELAP result. The time of scram also occur six seconds earlier in TRIP calculation compared to the RELAP. This exercise has also shown a proof of principle that coupling 3D

Assessment and improvement of condensation model in RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Rho, Hui Cheon; Choi, Kee Yong; Park, Hyeon Sik; Kim, Sang Jae [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Lee, Sang Il [Korea Power Engineering Co., Inc., Seoul (Korea, Republic of)

1997-07-15

The objective of this research is to remove the uncertainty of the condensation model through the assessment and improvement of the various heat transfer correlations used in the RELAP5/MOD3 code. The condensation model of the standard RELAP5/MOD3 code is systematically arranged and analyzed. A condensation heat transfer database is constructed from the previous experimental data on various condensation phenomena. Based on the constructed database, the condensation models in the code are assessed and improved. An experiment on the reflux condensation in a tube of steam generator in the presence of noncondensable gases is planned to acquire the experimental data.

Peer review of RELAP5/MOD3 documentation

International Nuclear Information System (INIS)

Craddick, W.G.

1994-01-01

A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Vol. III, Developmental Assessment Problems, and Vol. IV, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better discussion of discrepancies between the code and experimental data, and better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Vol. VI, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax's Equivalence Theorem

Application of a generalized interface module to the coupling of PARCS with both RELAP5 and TRAC-M

Energy Technology Data Exchange (ETDEWEB)

Barber, D.A.; Wang, W. [SCIENTECH, Inc. (United States); Miller, R.M.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States); Joo, H.G. [Korean Atomic Energy Research Inst., Seoul (Korea, Republic of); Mousseau, V.A. [Los Alamos National Lab., NM (United States); Ebert, D.E. [Nuclear Regulatory Commission, Washington, DC (United States)

1999-04-01

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 Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem.

A generic semi-implicit coupling methodology for use in RELAP5-3D(c)

International Nuclear Information System (INIS)

Weaver, W.L.; Tomlinson, E.T.; Aumiller, D.L.

2002-01-01

A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3D (c) computer program. This methodology allows RELAP5-3D (c) to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered and may use different numbers of conservation equations to model fluid flow in their respective solution domains. The methodology was demonstrated using a test case in which the test geometry was divided into two parts, each of which was solved as a RELAP5-3D (c) simulation. This test problem exercised all of the semi-implicit coupling features that were implemented in RELAP5-3D (c) The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process

Assessment of RELAP5/MOD3.2.2γ against flooding database in horizontal-to-inclined pipes

International Nuclear Information System (INIS)

Kim, Hyoung Tae; No, Hee Cheon

2001-01-01

A total of 356 experimental data for the onset of flooding are compiled for the data bank and used for the assessment of RELAP5/MOD3.2.2γ predictions of Counter-Current Flow Limitation (CCFL) in horizontal-to-inclined pipes simulating a PWR hot leg. The predictions of the flooding gas velocity in the database are known to be largely dependent on the horizontal pipe length-to-diameter ratio (L/D). RELAP5 calculations are compared with the experimental data where L/D is varied within the range of database. The present input model used for the simulation of CCFL is validated to reasonably calculate the gradient of water level in the horizontal pipes connected with the inclined volumes. RELAP5 calculations show that the RELAP5 predicts the flooding points qualitatively well but higher gas flow rate is required to initiate the flooding compared with the experimental data if the L/D is as low that of the hot legs of typical PWRs. Standard RELAP5 code is modified to apply the user specified CCFL curve not only to veritical volumes but also to the horizontal volumes. The calculation value by the modified version lies well on the applied CCFL curve even if flooding occurs at lower gas velocity thatn predicted by the CCFL curve in standard RELAP5

Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

International Nuclear Information System (INIS)

Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

2013-01-01

Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

Analysis of OECD/CSNI ISP-42 phase A PANDA experiment using coupled code R5G (RELAP5-GOTHIC)

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Grgic, D.; Bajs, T.

2010-01-01

In the paper, the results of the analysis of OECD/CSNI ISP-42 Phase A experiment at PANDA facility using stand-alone codes RELAP5/mod3.3 and GOTHIC 7.2b as well as coupled code R5G (RELAP5/mod3.3-GOTHIC 7.2b) are presented. PANDA is a large-scale thermal-hydraulic test facility installed at PSI (Paul Scherrer Institute) in Switzerland. The OECD/CSNI ISP-42 test consists of six sequential phases (Phase A through F). The present work deals with the post-test calculation of the Phase A, including the break of the main steam line and the Passive Containment Cooling (PCC) System Start-Up. The objective of the test is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air. The calculation was performed using stand-alone RELAP5/mod3.3 and GOTHIC 7.2b models, and then the same calculation was performed using coupled code with RELAP5 being responsible for reactor part of the model and GOTHIC being responsible for containment part of the model. The prediction capability, running time and modeling aspects were discussed for all three cases. (authors)

Assessment of PWR Steam Generator modelling in RELAP5/MOD2

International Nuclear Information System (INIS)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3

Comparison between UMSICHT water hammer experiments and calculations using RELAP5/MOD3.3

International Nuclear Information System (INIS)

Messing, Ralf

2008-01-01

Water hammer phenomena regularly occur in piping systems of nuclear power plants, e.g. after the rapid closure of valves. As the loads are usually several times higher than during normal operation pipe rupture is possible and therefore the integrity of the nuclear power plant can be endangered. In recent years extensive studies have been performed to assess the capabilities of the widely-used transient thermohydraulic system codes like RELAP5 for the simulation of water hammers and pressure surges in piping systems. The parameters affecting the results of such simulations are on the one side related to the numerical method, in particular the grid step size Δx and time step size Δt, and on the other hand to the models for present physical effects like the fluid-structure-interaction (FSI), unsteady friction or the release of dissolved air. In many studies experimental data obtained at the Fraunhofer Institute UMSICHT at Oberhausen/Germany has been used for code validation. In a vast campaign reliable data has been measured under varying boundary conditions for pressure surges in a pipe after a fast valve closure. Details of the experimental set-up are described in /1/ and /2/. Tiselj and Petelin /3/ provided theoretical background on the properties of the numerical scheme used by RELAP5 and Tiselj and Cerne /4/ highlighted the behavior at very small time steps. Kaliatka and al. /5/ investigated the influence of the grid step size and the time step size in RELAP5 on pressure transients obtained in UMSICHT experiments. Neuhaus and Dudlik /2/,/6/ studied the effects of fluidstructure- interaction, unsteady friction and degassing of dissolved air in tape-water filled pipes on pressure surges. They compared their numerical results again with experimental data from the UMSICHT test facility and observed a significant impact of all three parameters on the pressure-surge amplitudes and frequency. Barten and al. /15/ performed calculations of experiment 329 of the UMSICHT

Modelling of the Rod Control System in the coupled code RELAP5-QUABOX/CUBBOX

International Nuclear Information System (INIS)

Bencik, V.; Feretic, D.; Grgic, D.

1999-01-01

There is a general agreement that for many light water reactor transient calculations, it is necessary to use a multidimensional neutron kinetics model coupled to sophisticated thermal-hydraulic models in order to obtain satisfactory results. These calculations are needed for a variety of applications for licensing safety analyses, probabilistic risk assessment, operational support, and training. At FER, Zagreb, a coupling of 3D neutronics code QUABOX/CUBBOX and system code RELAP5 was performed. In the paper the Rod Control System model in the RELAP5 part of the coupled code is presented. A first testing of the model was performed by calculation of reactor trip from full power for NPP Krsko. Results of 3D neutronics calculation obtained by coupled code QUABOX/CUBBOX were compared with point kinetics calculation performed with RELAP5 stand alone code.(author)

Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

International Nuclear Information System (INIS)

Soares, Alexandre de Souza

2014-01-01

The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

Simulation of Spray Injection in the Pressurizer Using RELAP5

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S. Dibyo

2017-08-01

Full Text Available A modeling research using Relap5 to assess the pressurizer of a pressurized water reactor(PWR power plant has been performed. The heater and water injection systems in the pressurizer system of the PWRare of greatimportance for system pressure control.The heater is designed to increase the pressure while the water sprayer injection is to perform depressurization. Most of studies conducted in the past mainly focused on determining the effects of nozzle spray design and droplet size using testing loops. The purpose of this simulation is to analyze the spray injection flow rate against the pressure characteristics of the pressurizer using RELAP5. Through this approach, the optimum injection flow rate of full scale plant pressurizer can be analyzed. The parameters investigated are pressure and temperature.In RELAP5, the pressurizer tank wasmodeled with six volume nodes and the heater was modeled by using heat structure. In the model, the sprayer takes water from the cold leg to inject it into the top of tank region.The resultsshowedthat the mass flow of about 4 kg/s is the mosteffectivevalueto limit pressure in the pressurizer to below 15.7 MPa. However, the flow rates of 8 kg/s and more cause overpressure. This simulation is usefulto complement the data related to the water flow rate injection systems of the pressurizer. Normal 0 false false false EN-US X-NONE X-NONE Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

International Nuclear Information System (INIS)

Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

1989-01-01

In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

Benchmark evaluation of the RELAP code to calculate boiling in narrow channels

International Nuclear Information System (INIS)

Kunze, J.F.; Loyalka, S.K.; McKibben, J.C.; Hultsch, R.; Oladiran, O.

1990-01-01

The RELAP code has been tested with benchmark experiments (such as the loss-of-fluid test experiments at the Idaho National Engineering Laboratory) at high pressures and temperatures characteristic of those encountered in loss-of-coolant accidents (LOCAs) in commercial light water power reactors. Application of RELAP to the LOCA analysis of a low pressure (< 7 atm) and low temperature (< 100 degree C), plate-type research reactor, such as the University of Missouri Research Reactor (MURR), the high-flux breeder reactor, high-flux isotope reactor, and Advanced Test Reactor, requires resolution of questions involving overextrapolation to very low pressures and low temperatures, and calculations of the pulsed boiling/reflood conditions in the narrow rectangular cross-section channels (typically 2 mm thick) of the plate fuel elements. The practical concern of this problem is that plate fuel temperatures predicted by RELAP5 (MOD2, version 3) during the pulsed boiling period can reach high enough temperatures to cause plate (clad) weakening, though not melting. Since an experimental benchmark of RELAP under such LOCA conditions is not available and since such conditions present substantial challenges to the code, it is important to verify the code predictions. The comparison of the pulsed boiling experiments with the RELAP calculations involves both visual observations of void fraction versus time and measurements of temperatures near the fuel plate surface

RELAP5 simulations of critical break experiments in the RD-14 test facility

Energy Technology Data Exchange (ETDEWEB)

Kim, I-G; Cho, Y-J; Lee, S [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1996-12-31

RELAP5/MOD3 simulations of critical break tests in the RD-14 facility, modelling a loss of coolant in a CANDU reactor, were compared to the experimental results, and to CATHENA simulations of the early stage of the test. The RELAP5/MOD3 predicted thermal hydraulic behaviour reasonably well, but some discrepancies were observed after emergency cooling injection (ECI). Pressure differences between headers govern flow through the heated sections, particularly after ECI, and there is much uncertainty in the header pressures; further work is therefore recommended. 6 refs., 3 figs.

SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

International Nuclear Information System (INIS)

Achilli, A.; Congiu, C.; Ferri, R.; Bianchi, F.; Meloni, P.; Grgic, D.; Dzodzo, M.

2012-01-01

An Italian MSE R and D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

Directory of Open Access Journals (Sweden)

Andrea Achilli

2012-01-01

Full Text Available An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

Analysis with SCDAP/RELAP5 of reflooding of an overheated core in Forsmark 3 BWR after loss of electric power

International Nuclear Information System (INIS)

Nilsson, Lars.

1993-01-01

In foregoing SIK-2.2 project two severe accident sequences for Forsmark 3 comprising loss of electric power (Total Blackout, TB) without recovery actions (e.g. emergency cooling) were analysed with SCDAP/RELAP5. Code version MOD2.5/V3f was applied and a single core channel input model was used. Present analysis was done with the same initiating events as in SIK-2.2, but assuming recovery of auxiliary feed water and/or emergency core cooling systems to take place after heat-up of the core, in compliance with the plans of the SIK-2.3 project. A new test version of the SCDAP/RELAP5 code, version MOD3/V7af, was used here. The geometrical input model was extended from having one, into five parallel core zones, still with ten axial core sub volumes. Calculations were performed for three TB cases, one without cooling recovery, and two with injection of cold water at a rate of 45 kg/s, beginning when the maximum cladding temperature has reached 1522 K. The results show that a considerably more efficient cooling was obtained spraying the water to the top of the core, compared to the case where the cooling water was introduced into the downcomer, leading to bottom reflooding

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2015-01-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm 2 -rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

Framatome's experience in implementing and runnig RELAP5 MOD1

Energy Technology Data Exchange (ETDEWEB)

Truong, T.X.; Rousset, P.

1984-12-01

Implementation of RELAP MOD1 on Framatome's computer began in 1982, when we were working with the electrical power research institute on a safety and relief valve test program. It has been carried out in two stages: a first implementation on the CISI CYBER 740 computer; a transfer of files and a second on implemented our own CYBER 835 computer. The RELAP5 version currently implemented and used in Framatome is the cycle 19 standard version, no modification has been made yet, though some changes in data output files are intended.

Sensitivity Studies for Main Steam Line Break Exercises 2 and 3 with RELAP5/PANBOX

International Nuclear Information System (INIS)

Boeer, Rainer; Knoll, Alfred

2003-01-01

This paper presents and discusses results obtained with the nuclear plant safety analysis code system RELAP5/PANBOX (R/P/C) for the return-to-power scenario of exercises 2 and 3 of the Organization for Economic Cooperation and Development/Nuclear Energy Agency Main Steam Line Break (MSLB) Benchmark. Both the external and internal coupling options of R/P/C have been considered for exercise 3; i.e., the COBRA module of PANBOX was used to calculate the core thermal hydraulics in the external coupling option, whereas the core thermal hydraulics of RELAP5 was used in the internal coupling option. For the representation of thermal-hydraulic channels, a fine channel geometry based on the 177 fuel assemblies was selected for the external coupling option, and a coarse channel geometry based on 19 coarse channels has been investigated for the internal coupling option. The comparison of the results shows very good agreement of important core parameters between the considered coupling variants. Both exercises 2 and 3 have been investigated with respect to local safety parameters like fuel centerline temperatures and minimum departure from nucleate boiling ratios using the on-line hot subchannel analysis capability of R/P/C in the external coupling option. The results show that both quantities are far from the safety-related limits.The benchmark demonstrates, that R/P/C - as part of the integrated CASCADE-3D core analysis system of Framatome ANP GmbH - has proven to be a powerful tool for detailed analyses of an MSLB accident

Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

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Po Hu

2014-01-01

Full Text Available The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.

Analysis of panthers full-scale heat transfer tests with RELAP5

International Nuclear Information System (INIS)

Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

1996-01-01

The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric's (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit

Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

Energy Technology Data Exchange (ETDEWEB)

Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

2010-10-15

This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

Sensitive analysis and modifications to reflood-related constitutive models of RELAP5

International Nuclear Information System (INIS)

Li Dong; Liu Xiaojing; Yang Yanhua

2014-01-01

Previous system code calculation reveals that the cladding temperature is underestimated and quench front appears too early during reflood process. To find out the parameters shows important effect on the results, sensitive analysis is performed on parameters of constitutive physical models. Based on the phenomenological and theoretical analysis, four parameters are selected: wall to vapor film boiling heat transfer coefficient, wall to liquid film boiling heat transfer coefficient, dry wall interfacial friction coefficient and minimum droplet diameter. In order to improve the reflood simulation ability of RELAP5 code, the film boiling heat transfer model and dry wall interfacial friction model which are corresponding models of those influential parameters are studied. Modifications have been made and installed into RELAP5 code. Six tests of FEBA are simulated by RELAP5 to study the predictability of reflood-related physical models. A dispersed flow film boiling heat transfer (DFFB) model is applied when void fraction is above 0.9. And a factor is multiplied to the post-CHF drag coefficient to fit the experiment better. Finally, the six FEBA tests are calculated again so as to assess the modifications. Better results are obtained which prove the advantage of the modified models. (author)

ISP28. Calculation of PHEBUS test B9+ using SCDAP/RELAP5

International Nuclear Information System (INIS)

Eriksson, John.

1994-01-01

The SFD (Severe Fuel Damage) test B9+ carried out on the French PHEBUS facility of CEA in Cadarache provided the basis of the ISP28 (International standard Problem no.28). The main objective of the standard problem was to contribute to the assessments of codes used for calculation of phenomena during SFD accidents in a PWRs. ISP28 was organized partly as an open exercise in the sense that measured thermal-hydraulic conditions were available but in principle the ISP was a blind SFD exercise in which the degradation of the test bundle should be predicted. The Swedish participation in ISP28 relied on the SCDAP/RELAP5 best-estimate code which has been developed for calculations of fuel damage scenarios. As the calculation showed the code could successfully predict the main phenomena observed in the experiment. Contrary to the experiment, however, no cladding rupture occurred in the calculation. The main reason for this was the slightly too high thermal conductivity used for the thermal shield which caused not high enough peak temperatures

Modelling of QUENCH-03 and QUENCH-06 Experiments Using RELAP/SCDAPSIM and ASTEC Codes

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Tadas Kaliatka

2014-01-01

Full Text Available To prevent total meltdown of the uncovered and overheated core, the reflooding with water is a necessary accident management measure. Because these actions lead to the generation of hydrogen, which can cause further problems, the related phenomena are investigated performing experiments and computer simulations. In this paper, for the experiments of loss of coolant accidents, performed in Forschungszentrum Karlsruhe, QUENCH-03 and QUENCH-06 are modelled using RELAP5/SCDAPSIM and ASTEC codes. The performed benchmark allowed analysing different modelling features. The recommendations for the model development are presented.

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

RELAP5 progress summary: simulation of semiscale isothermal blowdown (Test S-01-4A)

International Nuclear Information System (INIS)

Kuo, H.H.; Wagner, R.J.; Carlson, K.E.; Kiser, D.M.; Trapp, J.A.; Ransom, V.H.

1978-07-01

The RELAP5/MOD''O'' LOCA analysis code has been applied to Simulation of the Semiscale Isothermal Blowdown Test (S-01-4A) from initiation to 60 seconds. Subcooled ECC injection was simulated from 23 seconds until accumulator emptying. The calculated results are in very good agreement with the experimental data. This is the first full system application of the RELAP5 code and only the pressurizer surge line resistance was modified to achieve the results reported. An analysis of the code execution time using a time-step statistical edit is included

Development, assessment, licensing, and application of RELAP5YA at Yankee Atomic Electric Company

International Nuclear Information System (INIS)

Husain, A.

1984-01-01

Since 1975, Yankee Atomic Electric Company (YAEC) has been licensed to perform LOCA analyses (large and small breaks) for PWRs. The methods are based upon the WREM package, comprised of RELAP4-MOD 3, RELAP4-FLOOD, and TOODEE-2EM. Significant efforts were expended to compare WREM models against Appendix K criteria, and to incorporate changes for compliance with the criteria and for improved representation of LOCA phenomena. /SUP 2,3,4/ This internalized capability has been used extensively for licensing the Yankee and Maine Yankee plants, and for timely resolution of LOCA issues which frequently arose. The value of this internalized analysis capability was well recognized by the Yankee management. In 1979, Yankee embarked upon another LOCA Methods Development Program. One objective was to remove extra conservatisms inherent in older codes to provide additional operational flexibility for our plants. Also, we needed a code which could be used in a production mode to provide realistic response for off-normal conditions to be used in operator training and emergency guideline assessment. This program resulted in the development of RELAP5YA. The development, assessment, licensing, and application of RELAP5YA at YAEC is briefly described in this paper

Preliminary results of L2 facility dynamics with RELAP5

International Nuclear Information System (INIS)

Mezio, F.; Giménez, M

2011-01-01

Conclusions: • We are working in the loop model; • RELAP5 may have some severe problems predicting the phenomena under study: –Some Correlations may be out of applicability domain; –Its numeric scheme is a diffusive one, which tends to artificially stabilize the system. But this wasn’t observed

Conversion tool for the LWR transient analysis code RELAP5 from the CDC version to the FACOM version

International Nuclear Information System (INIS)

Shinozawa, Naohisa; Fujisaki, Masahide; Makino, Mitsuhiro; Kondou, Kazuya; Ishiguro, Misako

1987-01-01

The LWR transient analysis code RELAP5 has been developed on the CDC-CYBER 176 at Idaho National Engineering Laboratory (INEL), the RELAP5 code has been often updated in order to extend the analyzing model and correct the errors. At Japan Atomic Energy Research Institute the code has been converted from the CDC version to the FACOM version and the converted code has been used. The conversion is the task which consumes a lot of time, because the code is large and there is the difference between CDC's machines and FACOM's ones. In order to convert the RELAP5 code automatically, the software tool has been developed. By using this tools the efficiency for converting the RELAP5 code has been improved. Productivity of the conversion is increased about 2.0 to 2.6 times by the tools in comparison with in manual. The procedure of conversion by using the tools and the option parameters of each tool are described. (author)

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

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Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Analysis of the preliminary trajectory of emergency venting of the nuclear power plant of Laguna Verde using RELAP5; Analisis de la trayectoria preliminar del venteo de emergencia de la Central Laguna Verde mediante RELAP5

Energy Technology Data Exchange (ETDEWEB)

Cecenas F, M.; Jimenez S, R.; Ovando C, R. [Instituto Nacional de Electricidad y Energias Limpias, Reforma No. 113, Col. Palmira, 42490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R. N., E-mail: mcf@iie.org.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km. 7.5, 94271 Veracruz (Mexico)

2016-09-15

For a commercial nuclear plant, the availability of a vent line to the atmosphere is an improvement to achieve the prevention and mitigation of the consequences of a severe accident. The importance of this system received greater attention after the Fukushima accident in 2011. Subsequently, in 2012 and 2013, the United States Nuclear Regulatory Commission issued documents stating that the venting must be able to dislodge 1% of the rated thermal power of the core without over pressurizing the primary container. To analyze the venting of the nuclear power plant of Laguna Verde, the line and the primary container are modeled using the thermo-hydraulic code RELAP5, simulating a release of 1% of the nominal licensed power to the container in the form of saturated steam. The vent has no problem to evacuate the energy and manages to keep the container without exceeding its design limit, and the highest percentage of thermal power that can channel the vent to the outside is approximately 3%. A sensitivity analysis increasing the diameter of the line to 14 inches allows increasing in 10% the percentage of power that can be vented to the outside without problem for the containment. (Author)

Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

1991-01-01

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 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

Assessment of RELAP5/MOD3.1 using LOFT L2-3 experiment data

International Nuclear Information System (INIS)

Lee, Sang Yong; Ban, Chang Hwan; Chung, Bob Dong

1994-06-01

The capability of RELAP5/MOD3.1 to predict overall LOCA thermal hydraulic phenomena was assessed utilizing the data of LOFT L2-3 experiment. Loop behaviors such as mass flow rate, water density, momentum flux, and the heating-up and rewetting of the fuel rod cladding during blowdown were well calculated. Reflood heat-up of the fuel rod cladding at the high power region of the core was reasonably predicted. But in the upper part of the core, cladding heat-up was calculated incorrectly since present code has no capability to calculate the top-down quenching which of highly multi-dimensional behavior. (Author) 10 refs., 46 figs., 2 tabs

RELAP4/MOD-5-CEA pump coastdown experiment simulation

International Nuclear Information System (INIS)

Borges, R.C.; Freitas, R.L.

1988-07-01

Since is important the theoretical-experimental comparison to evaluate the computer codes, these paper presents the simulation with RELAP4/MOD5 Code of a loss of power energy in the pump of the ''Circuito Experimental de Agua-CEA''. From the results attained, the existing models in the Code showed to be very satisfatory quantitative and qualitative behavior of the attained experimental results. (author) [pt

Blind-blind prediction by RELAP5/MOD1 for a 0.1% very small cold-leg break experiment at ROSA-IV large-scale test facility

International Nuclear Information System (INIS)

Koizumi, Y.; Kumamaru, H.; Kukita, Y.; Kawaji, M.; Osakabe, M.; Schultz, R.R.; Tanaka, M.; Tasaka, K.

1986-01-01

The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for the analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily

Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

International Nuclear Information System (INIS)

Loomis, G.G.; Streit, J.E.

1985-11-01

Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature [15.6 MPa (2262 psia) system pressure; 37 K (67 0 F) core differential temperature; 595 K(610 0 F) hot leg fluid temperature]. From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

International Nuclear Information System (INIS)

Yang, C.-Y.; Liang, Thomas K.S.; Pei, B.S.; Shih, C.K.; Chiang, S.C.; Wang, L.C.

2009-01-01

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its' own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.

RELAP5-3D Resolution of Known Restart/Backup Issues

Energy Technology Data Exchange (ETDEWEB)

Mesina, George L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Anderson, Nolan A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-12-01

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

International Nuclear Information System (INIS)

Szczurek, J.

1995-01-01

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

1995-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

1996-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Application of the Relap5-3D to phase 1 and 3 of the OECD-CSNI/NSC PWR MSLB benchmark related to TMI-1

International Nuclear Information System (INIS)

D'Auria, F.; Galassi, G.; Spadoni, A.; Hassan, Y.

2001-01-01

The Relap5-3D, the latest in the series of the Relap5 code, distinguishes from the previous versions by the fully integrated, multi-dimensional thermalhydraulic and kinetic modeling capability. It has been applied to Phase I and III of OECD-CSNI/ NSC PWR MSLB Benchmark adopting the same thermalhydraulic input deck already used with Relap5/Parcs and Relap5/Quabbox coupled codes during the previous MSLB analysis. The OECD jointly with the US NRC proposed the PWR MSLB Benchmark in order to gather a common understanding about the coupling between thermal hydraulics and neutronics, and evaluating the behavior of this transient with different coupled codes, giving emphasis to the 3-D modeling. This paper deals with the application of Relap5-3D code to phase I and III of the PWR MSLB Benchmark. The Relap5-3D is a thermal hydraulics-neutronics internally coupled code, the thermal hydraulics module is the INEEL version of Relap and the neutronics module is derived from NESTLE multi-dimension kinetics code. (author)

Assessment of RELAP5/MOD2 against critical flow data from Marviken tests JIT 11 and CFT 21

International Nuclear Information System (INIS)

Rosdahl, O.; Caraher, D.

1986-09-01

RELAP5/MOD2 simulations of the critical flow of saturated steam are reported together with simulations of the critical flow of subcooled liquid and a low quality two-phase mixture. The experiments which were simulated used nozzle diameters of 0.3 m and 0.5 m. RELAP5 overpredicted the experimental flow rates by 10 to 25% unless discharge coefficients were applied

Posttest analysis of international standard problem 10 using RELAP4/MOD7

International Nuclear Information System (INIS)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.; Behling, S.R.

1981-01-01

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of a hypothetical loss-of-coolant accident

RELAP5 kinetics model development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Judd, J.L.; Terry, W.K.

1990-01-01

A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

Standard model of WWER-440 fuel rod for Transuranus and its application for RELAP5 hot channel validation

International Nuclear Information System (INIS)

Hatala, B.; Cvan, M.

2001-01-01

Within the PECO European Commission project of 'Extension of the validation matrix of the TRANSURANUS code' is developed a generic model of WWER-440 fuel rod. The model is intended to be applied for both realistic and licensing, conservative analysis. For such an application the TRANSURANUS code would be complementary tool to generally used system codes, e.g. RELAP5, providing realistic, more detailed insight into processes and safety criteria, relevant to the fuel rod. The paper presents general description of the model for TRANSURANUS code, brief discussion of approaches used in TRANSURANUS and RELAP5 code safety analysis, accompanied with information about RELAP5 model (whole scope unit model, used for licensing analysis). The existing model for RELAP5 code for WWER-440/V-213 Bohunice V2 unit is checked and modified in hot channel part to allow transparent comparison with the TRANSURANUS code. The results from comparison calculations of the both codes are presented for fresh fuel and quasi steady state scenario and are in good agreement, almost identical. These results might be used as a basis for transient analysis

PHISICS/RELAP5-3D Adaptive Time-Step Method Demonstrated for the HTTR LOFC#1 Simulation

Energy Technology Data Exchange (ETDEWEB)

Baker, Robin Ivey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Balestra, Paolo [Univ. of Rome (Italy); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-05-01

A collaborative effort between Japan Atomic Energy Agency (JAEA) and Idaho National Laboratory (INL) as part of the Civil Nuclear Energy Working Group is underway to model the high temperature engineering test reactor (HTTR) loss of forced cooling (LOFC) transient that was performed in December 2010. The coupled version of RELAP5-3D, a thermal fluids code, and PHISICS, a neutronics code, were used to model the transient. The focus of this report is to summarize the changes made to the PHISICS-RELAP5-3D code for implementing an adaptive time step methodology into the code for the first time, and to test it using the full HTTR PHISICS/RELAP5-3D model developed by JAEA and INL and the LOFC simulation. Various adaptive schemes are available based on flux or power convergence criteria that allow significantly larger time steps to be taken by the neutronics module. The report includes a description of the HTTR and the associated PHISICS/RELAP5-3D model test results as well as the University of Rome sub-contractor report documenting the adaptive time step theory and methodology implemented in PHISICS/RELAP5-3D. Two versions of the HTTR model were tested using 8 and 26 energy groups. It was found that most of the new adaptive methods lead to significant improvements in the LOFC simulation time required without significant accuracy penalties in the prediction of the fission power and the fuel temperature. In the best performing 8 group model scenarios, a LOFC simulation of 20 hours could be completed in real-time, or even less than real-time, compared with the previous version of the code that completed the same transient 3-8 times slower than real-time. A few of the user choice combinations between the methodologies available and the tolerance settings did however result in unacceptably high errors or insignificant gains in simulation time. The study is concluded with recommendations on which methods to use for this HTTR model. An important caveat is that these findings

Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

1996-02-01

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 RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-09-01

In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

International Nuclear Information System (INIS)

Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu

1985-03-01

The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)

Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

International Nuclear Information System (INIS)

Rizzo, G.; Vella, G.

2007-01-01

The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

Prediction of loop seal formation and clearing during small break loss of coolant accident

International Nuclear Information System (INIS)

Lee, Suk Ho; Kim, Hho Jung

1992-01-01

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD2 and /MOD3 codes with the test of SB-CL-18 of the LSTF(Large Scale Test Facility). The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery includeing the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/ MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in the base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs with the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing. (Author)

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Assessment of the Radiation Enclosure Models in SPACE and RELAP5 with GOTA Test 27

Energy Technology Data Exchange (ETDEWEB)

Lee, T. B.; Lee, G. W.; Choi, T. S. [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

SPACE (Safety and Performance Analysis Code) for nuclear power plant has been developed to calculate the transient thermal-hydraulic response of PWRs that can contain multiple types of fluids. Without explaining 3-D effects such as the change of fuel rod/guide tube thermal behavior as a result of the radiation heat transfer, the 1-D code could predict an unrealistically high peak clad temperature. A useful function to simulate the wall-to-wall radiation heat transfer is implemented in the SPACE and RELAP5 codes. This paper discusses the assessment results of the radiation enclosure model of SPACE and RELAP5. The capability of handling wall-to-wall radiation problem of the SPACE and the RELAP5 codes has been evaluated using the experimental data from the GOTA test facility. At the top of the bundle, the maximum errors of SPACE and RELAP5 are less than 1.6% and 2.3%, respectively. As noted, there is a small discrepancy between the calculated results and experimental data except for the predictions near the top of the test section. The SPACE code is based on the version 2.16 distributed by KHNP. In order to perform the simulation of the GOTA test 27, it was necessary to modify the SPACE code. There was the subroutine for an input process corresponding to the radiation model, the inp{sub c}heck function of the RadEncData Class, contained in a vulnerable algorithm to figure out the reciprocity rule of the view factor.

Analysis of steam generator behaviour in nuclear power plant with computer program RELAP5; Analiza delovanja uparjalnika jedrske elektrarne s programom RELAP5

Energy Technology Data Exchange (ETDEWEB)

Zeljko, M; Gregoric, M; Peterlin, G [Institut ' Jozef Stefan' , Ljubljana (Yugoslavia)

1983-07-01

Analyses of nuclear power plant behaviour are made with large computer programs. We used RELAP5/MOD1/CYCLE001, which was developed in Idaho National Engineering laboratory. Input model was prepared to analyze transients in steam generator of NPP Krsko. We found out that this version had a lot of faults so we intend to implement a new cycle. First experience shows the whole complexity of such analysis from technical and economical viewpoints. (author)

Small break LOCA RELAP5/MOD3 uncertainty quantification: Bias and uncertainty evaluation for important phenomena

International Nuclear Information System (INIS)

Ortiz, M.G.; Ghan, L.S.; Vogl, J.

1991-01-01

The Nuclear Regulatory Commission (NRC) revised the Emergency Core Cooling System (ECCS) licensing rule to allow the use of Best Estimate (BE) computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability and Uncertainty (CSAU) to evaluate BE code uncertainties. The CSAU methodology was demonstrated with a specific application to a pressurized water reactor (PWR), experiencing a postulated large break loss-of-coolant accident (LBLOCA). The current work is part of an effort to adapt and demonstrate the CSAU methodology to a small break (SB) LOCA in a PWR of B and W design using RELAP5/MOD3 as the simulation tool. The subject of this paper is the Assessment and Ranging of Parameters (Element 2 of the CSAU methodology), which determines the contribution to uncertainty of specific models in the code

Assessment of RELAP5/MOD2 against a natural circulation experiment in Nuclear Power Plant Borssele

International Nuclear Information System (INIS)

Winters, L.

1993-07-01

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 circulation experiment as conducted at the Borssele Nuclear Power Plant. The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele

Coupled RELAP5/GOTHIC model for IRIS SBLOCA analysis

International Nuclear Information System (INIS)

Grgic, D.; Cavlina, N.; Bajs, T.; Oriani, L.; Conway, L. E.

2004-01-01

. However, the time that would be needed to develop new codes where all required models are properly incorporated, to build user experience, and to qualify the code would be prohibitively long and expensive. Therefore, the coupling of the thermal-hydraulic and containment codes can be an interesting approach and compromise between two modeling strategies mentioned above. Separate, existing computer codes can be coupled providing new capabilities without spending too much time in development and with possibility to use existing experience and perform code verification and validation only for the coupling portion of the new code. This coupling is usually performed as an extension of the classical calculation approach and it is localized at the physical points where communication between system and containment exists. For IRIS, it was decided to develop an explicit coupling of RELAP5/mod3.3, and one of the earlier versions of the GOTHIC code available at University of Zagreb, GOTHIC 3.4e; thus taking advantage of the rather large experience base in the use of the RELAP5 and GOTHIC codes as well as knowledge of their internal structure The primary goal was to explore applicability of coupled code to safety analyses of the new reactor systems where the primary system and containment closely interact. The chosen coupling strategy is simple and basic operation of constituent codes and corresponding input data are unaffected by the coupling process. This paper describes the coupled code as well as the development of the preliminary IRIS SBLOCA evaluation model and its use. Also, a discussion on the verification and validation of this methodology is provided.(author)

Assessment of RELAP5/MOD3.2 with condensation experiment in the presence of noncondensables in a vertical tube

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, and simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube. 10 refs., 6 figs. (Author)

Assessment of RELAP5/MOD3.2 with condensation experiment in the presence of noncondensables in a vertical tube

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, and simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube. 10 refs., 6 figs. (Author)

Some Comments on the Behavior of the RELAP5 Numerical Scheme at Very Small Time Steps

International Nuclear Information System (INIS)

Tiselj, Iztok; Cerne, Gregor

2000-01-01

The behavior of the RELAP5 code at very short time steps is described, i.e., δt [approximately equal to] 0.01 δx/c. First, the property of the RELAP5 code to trace acoustic waves with 'almost' second-order accuracy is demonstrated. Quasi-second-order accuracy is usually achieved for acoustic waves at very short time steps but can never be achieved for the propagation of nonacoustic temperature and void fraction waves. While this feature may be beneficial for the simulations of fast transients describing pressure waves, it also has an adverse effect: The lack of numerical diffusion at very short time steps can cause typical second-order numerical oscillations near steep pressure jumps. This behavior explains why an automatic halving of the time step, which is used in RELAP5 when numerical difficulties are encountered, in some cases leads to the failure of the simulation.Second, the integration of the stiff interphase exchange terms in RELAP5 is studied. For transients with flashing and/or rapid condensation as the main phenomena, results strongly depend on the time step used. Poor accuracy is achieved with 'normal' time steps (δt [approximately equal to] δx/v) because of the very short characteristic timescale of the interphase mass and heat transfer sources. In such cases significantly different results are predicted with very short time steps because of the more accurate integration of the stiff interphase exchange terms

RELAP-7 Theory Manual

Energy Technology Data Exchange (ETDEWEB)

Berry, Ray Alden [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Peterson, John William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kadioglu, Samet Yucel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

This document summarizes the physical models and mathematical formulations used in the RELAP-7 code. In summary, the MOOSE based RELAP-7 code development is an ongoing effort. The MOOSE framework enables rapid development of the RELAP-7 code. The developmental efforts and results demonstrate that the RELAP-7 project is on a path to success. This theory manual documents the main features implemented into the RELAP-7 code. Because the code is an ongoing development effort, this RELAP-7 Theory Manual will evolve with periodic updates to keep it current with the state of the development, implementation, and model additions/revisions.

Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

Energy Technology Data Exchange (ETDEWEB)

Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

2015-12-15

Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

Assessment of core damage models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

International Nuclear Information System (INIS)

Coryell, E.W.

1991-01-01

The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

2010-07-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

RELAP5-3D developmental assessment: Comparison of version 4.2.1i on Linux and Windows

Energy Technology Data Exchange (ETDEWEB)

Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-06-01

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.2i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

RELAP5-3D Developmental Assessment. Comparison of Version 4.3.4i on Linux and Windows

International Nuclear Information System (INIS)

Bayless, Paul David

2015-01-01

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.3i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Study of the Relap5/mod3.2 wall heat flux partitioning model

International Nuclear Information System (INIS)

Hari, S.; Hassan, Y.A.

2001-01-01

The performance of the subcooled boiling model adapted in RELAP5/MOD3.2 computer code has been assessed in detail for low-pressure conditions and it has been found that the void fraction profile is under-predicted. In general, any subcooled boiling model is composed of individual sub-models that account for the different physical mechanism that govern the overall process, as the wall vapor generation, interfacial shear and condensation etc. The wall heat flux partitioning model is one of the important sub-models that is a constituent of any subcooled boiling model. The function of this model is to apportion the wall heat flux to the different components (as the single/two phase fluid or bubble), as the case may be, in a two-phase flow-boiling scenario adjacent to a heated wall. The ''pumping factor'' approach is generally followed by most of the wall heat flux partitioning models, for partitioning the wall heat flux. In this work, the wall heat flux partitioning model of RELAP5/MOD3.2 computer code is studied; in particular, the ''pumping factor'' formulation in the present code version is assessed for its performance under low-pressure conditions. In addition, three different ''pumping factor'' formulations available in the literature have been introduced into the RELAP5/MOD3.2 code. Simulations of two low-pressure subcooled flow boiling experiments were performed with the refined code versions to determine the appropriate pumping factor to be used under these conditions. (author)

Assessment of Flow Instability in Passive Auxiliary Feedwater System (PAFS) Using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cheon, Jong; Kim, Han-Gon [KHNP, Daejeon (Korea, Republic of)

2015-10-15

In this study, the occurrence possibility of both instabilities in PAFS is assessed with the best-estimate thermal hydraulic code, RELAP5. From the RELAP5 code analysis, the Ledinegg instability might not occur in PAFS. The DWO might occur in PAFS but the effect of the oscillation on the heat removal capacity of PAFS was not large. Therefore, it is concluded that PAFS is safe in terms of flow instabilities. Since PAFS is two-phase flow system, flow instabilities may occur. Flow instabilities may cause the severe deterioration of heat removal capability of PAFS due to the reduction of the condensate flow. For the reliable operation of PAFS, it is required to assess the flow instabilities in PAFS. The Ledinegg-type instability and the Density Wave Oscillation (DWO) are the representative static flow instability and the dynamic flow instability, respectively.

Preliminary Study of Steam Generator Water Level Tracking by Three Different Methods Using RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Park, Ki Moon; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-10-15

It has been identified in the previous works that the tracking of a steam generator (SG) water level is important. However, three different parameters can be used as an indicator of the SG water level. These parameters are: (1) SG downcomer collapsed water level, (2) water mass inventory and (3) pressure differential between upper and low tap of SG. Instead the SG water level is calculated by either SG downcomer collapsed water level or water mass inventory. However, the pressure differential measurement is the most widely used method for estimating the SG water level in the experiment as well as in the industry In this paper, therefore, three events are analyzed to perform sensitivity study of the SG water level calculation with RELAP5/MOD3 and evaluate SG level difference by three parameters. In this paper, three events are analyzed using the system analysis code (RELAP5/MOD3) to check for the consistency among the downcomer collapsed water level, mass inventory and the pressure differential measurement methods. This is to identify the sensitivity of the nuclear power plant accident response when one of the above three parameters is selected as the representative parameter of the steam generator water level. It is confirmed that mass inventory method is not affected by shrinking and swelling effect and the reactor trip time is significantly different among three parameters during TLOFW. In addition, level recovery rate is different when LOMF occurs. Thus, the SG level sensitivity of SG water level tracking method using three parameters has to be further studied not only for the steady-state operation but also for understanding the nuclear power plant response under various transient scenarios.

Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

International Nuclear Information System (INIS)

Roth, P.A.; Schultz, R.R.; Choi, C.J.

1992-07-01

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 depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests

The implementation of the CDC version of RELAP5/MOD1/019 on an IBM compatible computer system (AMDAHL 470/V8)

International Nuclear Information System (INIS)

Kolar, W.; Brewka, W.

1984-01-01

RELAP5/MOD1 is an advanced one-dimensional best estimate system code, which is used for safety analysis studies of nuclear pressurized water reactor systems and related integral and separate effect test facilities. The program predicts the system response for large break, small break LOCA and special transients. To a large extent RELAP5/MOD1 is written in Fortran, only a small part of the program is coded in CDC assembler. RELAP5/MOD1 was developed on the CDC CYBER 176 at INEL*. The code development team made use of CDC system programs like the CDC UPDATE facility and incorporated in the program special purpose software packages. The report describes the problems which have been encountered when implementing the CDC version of RELAP5/MOD1 on an IBM compatible computer systems (AMDAHL 470/V8)

Modernization and restructuring of realistic thermal hydraulic system analysis code, RELAP5/MOD3.3.1.2

International Nuclear Information System (INIS)

Chung, Bub Dong; Lee, Won Jae; Jeong, Jae Jun; Ha, Kwi Seok

1998-01-01

The code architecture entails the programming language and the code database. Various recent programming languages such as C, C ++ , Fortran 90, were considered as the candidate language for the modernization of RELAP5/MOD3.2.1.2. Among them, Fortran 90 was selected as a basic programming laguage for the modernization and restructuring of the code. Most of header file ( * .h) and equivalenced variables in RELAP5 have been replaced with members in the MODULE, which greatly enhance the code maintenance and readability. The FTB package is used for the dynamic memory management (DMM) of RELAP5. Although FTB DMM features are very successful, the use of FTB has been the obstacle in the maintenance of the code. It is difficult to understand and change the coding, and it requires a significant effort to find out index errors in large memory pools. With new features introduced in Fortran 90, it is possible to slove dynamic allocation problems within the standard features in an elegant, clear safe way. Each of FTB data blocks can be replaced by the suitably organized derived variables in MODULE and the standard DMM scheme. This DMM scheme provides the code flexibility which can save the memory requirements depending on the problem sizes without a extensive use of the complex FTB package. The current user's interface of the RELAP5 consists of a set of input file, output file, and restart/plot file. Many users complain that this interface is not user friendly. It was mainly caused by the text-oriented programming, namly console programming during the past many years. Now, windows programming has become popular in most areas of software development. Using this windows programming technique, the user friend freatures can be implemented. The Visual Fortran Quick Win run-time library helps to turn graphics programs into simple Windows applications. RELAP5 code has been re-compiled with the Quick Win feature, and the mask for user's dialog and graphical x-y plot were designed. This

A three-dimensional nodal neutron kinetics capability for relaps

International Nuclear Information System (INIS)

Judd, J.L.; Weaver, W.L.

1996-01-01

The incorporation of a three-dimensional neutron kinetics capability into the DOE version of the RELAP5/MOD3.2 reactor safety code is discussed. A brief discussion of the kinetics method is given along with a discussion of the cross section parameterization models available in RELAP5/MOD3.2. The RELAP5/MOD3.2 code is then used to perform calculations of the NEACRP rod ejection and rod withdrawal benchmarks, and results are presented

Assessment of 12 CHF prediction methods, for an axially non-uniform heat flux distribution, with the RELAP5 computer code

Energy Technology Data Exchange (ETDEWEB)

Ferrouk, M. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria)], E-mail: m_ferrouk@yahoo.fr; Aissani, S. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria); D' Auria, F.; DelNevo, A.; Salah, A. Bousbia [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (Italy)

2008-10-15

The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared.

Analysis, by RELAP5 code, of boron dilution phenomena in a mid-loop operation transient, performed in PKL III F2.1 RUN 1 test

International Nuclear Information System (INIS)

Mascari, F.; Vella, G.; Del Nevo, A.; D'Auria, F.

2007-01-01

The present paper deals with the post test analysis and accuracy quantification of the test PKL III F2.1 RUN 1 by RELAP5/Mod3.3 code performed in the framework of the international OECD/SETH PKL III Project. The PKL III is a full-height integral test facility (ITF) that models the entire primary system and most of the secondary system (except for turbine and condenser) of pressurized water reactor of KWU design of the 1300-MW (electric) class on a scale of 1:145. Detailed design was based to the largest possible extent on the specific data of Philippsburg nuclear power plant, unit 2. As for the test facilities of this size, the scaling concept aims to simulate overall thermal hydraulic behavior of the full-scale power plant [1]. The main purpose of the project is to investigate PWR safety issues related to boron dilution and in particular this experiment investigates (a) the boron dilution issue during mid-loop operation and shutdown conditions, and (b) assessing primary circuit accident management operations to prevent boron dilution as a consequence of loss of heat removal [2]. In this work the authors deal with a systematic procedure (developed at the university of Pisa) for code assessment and uncertainty qualification and its application to RELAP5 system code. It is used to evaluate the capability of RELAP5 to reproduce the thermal hydraulics of an inadvertent boron dilution event in a PWR. The quantitative analysis has been performed adopting the Fast Fourier Transform Based Method (FFTBM), which has the capability to quantify the errors in code predictions as compared to the measured experimental signal. (author)

A RELAP5 study to identify flow regime in natural circulation phenomenon

Energy Technology Data Exchange (ETDEWEB)

Sabundjian, Gaiane; Torres, Walmir M.; Macedo, Luiz A.; Mesquita, Roberto N.; Andrade, Delvonei A.; Umbehaun, Pedro E.; Conti, Thadeu N.; Masotti, Paulo H.F.; Belchior Junior, Antonio; Angelo, Gabriel, E-mail: gdjian@ipen.b, E-mail: umbehaun@ipen.b, E-mail: wmtorres@ipen.b, E-mail: tnconti@ipen.b, E-mail: rnavarro@ipen.b, E-mail: lamacedo@ipen.b, E-mail: pmasotti@ipen.b, E-mail: abelchior@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

There has been a crescent interest in the scientific community in the study of natural circulation phenomenon. New generation of compact nuclear reactors uses the natural circulation of the fluid as a system of cooling and of residual heat removal in case of accident or shutdown. The objective of this paper is to compare the flow patterns of experimental data and numerical simulation for the natural circulation phenomenon in two-phase flow regime. An experimental circuit built with glass tubes is used for the experiments. Thus, it allows the thermal hydraulic phenomena visualization. There is an electric heater as the heat source, a heat exchanger as the heat sink and an expansion tank to accommodate fluid density excursions. The circuit instrumentation consists of thermocouples and pressure meters to better keep track of the flow and heat transfer phenomena. Data acquisition is performed through a computer interface developed with LABVIEW. The characteristic of the regime is identified using photography techniques. Numerical modeling and simulation is done with the thermal hydraulic code RELAP5, which is widely used for this purpose. This numerical simulation is capable to reproduce some of the flow regimes which are present in the circuit for the natural circulation phenomenon. Comparison between experimental and numerical simulation is presented in this work. (author)

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

Analysis of reactivity transient for the DIDO type research reactors using RELAP5

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-Salah, A.; D'Auria, F.; Nabbi, R.

2005-01-01

Recent availability of high performance computers and computational methods together with the continuing increase in operational experience imposes revising some operational constrains and conservative safety margins. The application of Best-Estimate (BE) method constitutes a real necessity in the safety and design analysis and allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best-Estimate methods results in the mitigation of the constraining limits in design and operation. This paper presents the results of the application of the RELAP5/Mod3.3 system thermal-hydraulic code to the German FRJ-2 research reactor for a reactivity transient, which has been analyzed in the past using the verified system code CATHENA [1], [2], [3]. The work mainly aims checking the capability of RELAP5 [4] for research reactor transient analysis by the comparison of the results of the two codes and including modeling basis and analytical approaches. According to the existing references RELAP5 applications are concentrated on the transient analysis of nuclear power systems. The considered case consists of a simulation related to a hypothetical fast reactivity transient, which is assumed to be caused by the failure of one shutdown arm. The case has been chosen due to the importance of the models for the precise description of the complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For this purpose, the fuel element assembly was modeled in detail according to design data. The primary circuit was included in the whole model in order to consider the interaction with individual fuel elements with core. In general the results of the two codes are in agreement and comparable during the initial phase of the transient. After reaching the flow regime with fully developed nucleate boiling and two phase flow RELAP5 exhibits

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

International Nuclear Information System (INIS)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

Energy Technology Data Exchange (ETDEWEB)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs.

A study of core melting phenomena in reactor severe accident of PWR

Energy Technology Data Exchange (ETDEWEB)

Jeun, Gyoo Dong; Park, Shane; Kim, Jong Sun; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Jin Man [Korea Maritime Univ., Busan (Korea, Republic of)

2001-03-15

In the 4th year, SCDAP/RELAP5 best estimate input data obtained from the TMI-2 accident analysis were applied to the analysis of domestic nuclear power plant. Ulchin nuclear power plant unit 3, 4 were selected as reference plant and steam generator tube rupture, station blackout SCDAP/RELAP5 calculation were performed to verify the adequacy of the best estimate input parameters and the adequacy of related models. Also, System 80+ EVSE simulation was executed to study steam explosion phenomena in the reactor cavity and EVSE load test was performed on the simplified reactor cavity geometry using TRACER-II code.

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

Analysis of the preliminary trajectory of emergency venting of the nuclear power plant of Laguna Verde using RELAP5

International Nuclear Information System (INIS)

Cecenas F, M.; Jimenez S, R.; Ovando C, R.; Tijerina S, F.; Tapia M, R. N.

2016-09-01

For a commercial nuclear plant, the availability of a vent line to the atmosphere is an improvement to achieve the prevention and mitigation of the consequences of a severe accident. The importance of this system received greater attention after the Fukushima accident in 2011. Subsequently, in 2012 and 2013, the United States Nuclear Regulatory Commission issued documents stating that the venting must be able to dislodge 1% of the rated thermal power of the core without over pressurizing the primary container. To analyze the venting of the nuclear power plant of Laguna Verde, the line and the primary container are modeled using the thermo-hydraulic code RELAP5, simulating a release of 1% of the nominal licensed power to the container in the form of saturated steam. The vent has no problem to evacuate the energy and manages to keep the container without exceeding its design limit, and the highest percentage of thermal power that can channel the vent to the outside is approximately 3%. A sensitivity analysis increasing the diameter of the line to 14 inches allows increasing in 10% the percentage of power that can be vented to the outside without problem for the containment. (Author)

High Temperature Test Facility Preliminary RELAP5-3D Input Model Description

Energy Technology Data Exchange (ETDEWEB)

Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-12-01

A RELAP5-3D input model is being developed for the High Temperature Test Facility at Oregon State University. The current model is described in detail. Further refinements will be made to the model as final as-built drawings are released and when system characterization data are available for benchmarking the input model.

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

OpenAIRE

Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

2016-01-01

Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

Progress on B4C control rod modeling in RELAP/SCDAPSIM with application to quench and Phebus

International Nuclear Information System (INIS)

Kawahara, Keisuke; Hohorst, Judith K.; Allison, Chris M.

2014-01-01

The RELAP/SCDAPSIM code is designed to predict the behavior of reactor systems during normal and accident conditions. RELAP/SCDAPSIM/MOD3.5 is an experimental version of the code with the most advanced fuel and severe accident behavior models and correlations. It includes modeling improvements that were specifically added to support (a) the ongoing experimental severe accident programs in Europe and Japan and (b) the analysis and assessment activities related to the accident at the Fukushima Daiichi NPS. One of the improved models describes the behavior of cylindrical B 4 C control rods used in selected PWR designs and in integral experiments used to assess the heating and melting of PWR, BWR, and VVER assemblies. It replaces an older model that was originally developed by the US Nuclear Regulatory Commission in the mid- 1980's. It includes a combination of new and improved models and correlations to more accurately describe (a) eutectic reactions between Zircaloy, B 4 C, and stainless steel, (b) oxidation for B 4 C, Zircaloy, and stainless steel, and (c) the effects of the gap between the Zircaloy guide tube and the stainless steel sheath surrounding B 4 C pellets used in many control rod designs. This paper will discuss the development of the new model and validation of the model using the PHEBUS B 4 C test, FPT-3, and the KIT quench experiments with a central B 4 C control rod. (authors)

Investigation of modeling and simulation on a PWR power conversion system with RELAP5

International Nuclear Information System (INIS)

Rui Gao; Yang Yanhua; Lin Meng; Yuan Minghao; Xie Zhengrui

2007-01-01

Based on the power conversion system of nuclear and conventional islands of Dayabay nuclear power station, this paper models the thermal-hydraulic systems for PWR by using the best-estimate program, RELAP5. To simulate the full-scope power conversion system, not only the reactor coolant system (RCP) of nuclear island, but also the main steam system (VVP), turbine steam and drain system (GPV), bypass system (GCT), feedwater system (FW), condensate extraction system (CEX), moisture separator reheater system (GSS), turbine-driven feedwater pump (APP), low-pressure and high-pressure feedwater heater systems (ABP and AHP) of conventional island are considered and modeled. A comparison between the simulated results and the actual data of reactor under full-power demonstrates a fine match for Dayabay, and also manifests the feasibility in simulating full-scope power conversion system of PWR with RELAP5. (author)

Review of solution approach, methods, and recent results of the RELAP5 system code

International Nuclear Information System (INIS)

Trapp, J.A.; Ransom, V.H.

1983-01-01

The present RELAP5 code is based on a semi-implicit numerical scheme for the hydrodynamic model. The basic guidelines employed in the development of the semi-implicit numerical scheme are discussed and the numerical features of the scheme are illustrated by analysis for a simple, but analogous, single-equation model. The basic numerical scheme is recorded and results from several simulations are presented. The experimental results and code simulations are used in a complementary fashion to develop insights into nuclear-plant response that would not be obtained if either tool were used alone. Further analysis using the simple single-equation model is carried out to yield insights that are presently being used to implement a more-implicit multi-step scheme in the experimental version of RELAP5. The multi-step implicit scheme is also described

RELAP-7 Software Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Choi, Yong-Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support

2014-09-25

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

RELAP/REFLA (Mod 0): a system reflooding analysis computer program

International Nuclear Information System (INIS)

Fujiki, Kazuo; Murao, Yoshio; Shimooke, Takanori.

1981-03-01

A new computer code RELAP/REFLA has been developed, aiming at analyses of the core reflooding phenomena during the postulated loss-of-coolant accident of PWRs. The code was originated from the combination of two distinct codes, RELAP4-FLOOD and REFLA-1D. The characteristics of the code are: (1) Kinematical model based on the observation and analysis of quench experiments is used for the thermal-hydraulic analysis of reflooding core, (2) it has the capability to analyse the reflooding phenomena in an arbitrary type of PWR or experimental facility, including the system feedback effects, (3) the flow paths in the actual system are represented by the combination of 1-dimentional flow paths, and vapor-liquid equilibrium model is applied except the reflooding core. This report is a code manual of RELAP/REFLA (version Mod 0) and contains the descriptions of the basic models, basic equations, code structure and input format. The calculated results of two kinds of sample problems, i.e., reflooding problem on the 4 loop PWR and FLECHT-SET experiment, are also presented. Relatively close agreement between FLECHT-SET data and the calculated results was obtained for the lower portion of the core, but poor agreement for the temperature histories in the upper core and carryover ratio. Running speed and core memory size are almost equal to those of RELAP 4/Mod 3. (author)

Developmental assessment of RELAP5/MOD3 code against ROSA-IV/TPTF horizontal two-phase flow experiments

International Nuclear Information System (INIS)

Kukita, Yutaka; Asaka, Hideaki; Anoda, Yoshinari; Ishiguro, Misako; Tasaka, Kanji; Mimura, Yuichi; Nemoto, Toshiyuki.

1990-03-01

A developmental version of the RELAP5/Mod3 code (as of June 1989) was assessed for accuracy using experimental data taken for high-pressure (7MPa) steam-water two-phase flow in a large-diameter (0.18 m) horizontal-pipe test section of the ROSA-IV Two-Phase Flow Test Facility (TPTF). The agreement between the measured and calculated test section void fractions was much better than that for the previous generation of RELAP5 (MOD2). The improvement was achieved primarily due to the code changes with respect to the flow stratification criterion and interfacial-drag calculation scheme. (author)

RELAP5 Model Description and Validation for the BR2 Loss-of-Flow Experiments

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Koonen, E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-07-01

This paper presents a description of the RELAP5 model, the calibration method used to obtain the minor loss coefficients from the available hydraulic data and the LOFA simulation results compared to the 1963 experimental tests for HEU fuel.

Use of moving heat conductor mesh to perform reflood calculations with RELAP4/MOD6

International Nuclear Information System (INIS)

Fischer, S.R.; Ellis, L.V.; Chen, Y.S.

1979-01-01

RELAP4 is a computer code which can be used for the transient thermal hydraulic analysis of light water reactors and related systems. RELAP4/MOD6 includes many new analytical models which were developed primarily for the analysis of the reflood phase of a PWR loss-of-coolant accident (LOCA) transient. The key feature forming the basis for the MOD6 reflood calculation is a unique moving finite differenced heat conductor. The development and application of the moving heat conductor mesh for use in reflood analysis are described

RELAP5 assessment using semiscale SBLOCA test S-NH-1

International Nuclear Information System (INIS)

Lee, E.J.; Chung, B.D.; Kim, H.J.

1993-06-01

2-inch cold leg break test S-NH-1, conducted at the 1/1705 volume scaled facility Semiscale was analyzed using RELAP5/MOD2 Cycle 36.04 and MOD3 Version 5m5. Loss of HPIS was assumed, and reactor trip occurred on a low PZR pressure signal (13.1 MPa), and pumps began an unpowered coastdown on SI signal (12.5 MPa). The system was recovered by opening ADV's when the PCT became higher than 811 K. Accumulator was finally injected into the system when the primary system pressure was less than 4.0 MPa. The experiment was terminated when the pressure reached the LPIS actuation set point RELAP5/MOD2 analysis demonstrated its capability to predict, with a sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative points of view. Nevertheless, several differences were noted regarding the break flow rate and inventory distribution due to deficiencies in two-phase choked flow model, horizontal stratification interfacial drag, and a CCFL model. The main reason for the core to remain nearly fully covered with the liquid was the under-prediction of the break flow by the code. Several sensitivity calculations were tried using the MOD2 to improve the results by using the different options of break flow modeling (downward, homogeneous, and area increase). The break area compensating concept based on ''the integrated break flow matching'' gave the best results than downward junction and homogeneous options. And the MOD3 showed improvement in predicting a CCFL in SG and a heatup in the core

Validation of one-dimensional module of MARS 2.1 computer code by comparison with the RELAP5/MOD3.3 developmental assessment results

International Nuclear Information System (INIS)

Lee, Y. J.; Bae, S. W.; Chung, B. D.

2003-02-01

This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

RELAP5 two-phase fluid model and numerical scheme for economic LWR system simulation

International Nuclear Information System (INIS)

Ransom, V.H.; Wagner, R.J.; Trapp, J.A.

1981-01-01

The RELAP5 two-phase fluid model and the associated numerical scheme are summarized. The experience accrued in development of a fast running light water reactor system transient analysis code is reviewed and example of the code application are given

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

International Nuclear Information System (INIS)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-01-01

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-12-15

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Development, validation and application of multi-point kinetics model in RELAP5 for analysis of asymmetric nuclear transients

Energy Technology Data Exchange (ETDEWEB)

Pradhan, Santosh K., E-mail: santosh@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Obaidurrahman, K. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Iyer, Kannan N. [Department of Mechanical Engineering, IIT Bombay, Mumbai 400076 (India); Gaikwad, Avinash J. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India)

2016-04-15

Highlights: • A multi-point kinetics model is developed for RELAP5 system thermal hydraulics code. • Model is validated against extensive 3D kinetics code. • RELAP5 multi-point kinetics formulation is used to investigate critical break for LOCA in PHWR. - Abstract: Point kinetics approach in system code RELAP5 limits its use for many of the reactivity induced transients, which involve asymmetric core behaviour. Development of fully coupled 3D core kinetics code with system thermal-hydraulics is the ultimate requirement in this regard; however coupling and validation of 3D kinetics module with system code is cumbersome and it also requires access to source code. An intermediate approach with multi-point kinetics is appropriate and relatively easy to implement for analysis of several asymmetric transients for large cores. Multi-point kinetics formulation is based on dividing the entire core into several regions and solving ODEs describing kinetics in each region. These regions are interconnected by spatial coupling coefficients which are estimated from diffusion theory approximation. This model offers an advantage that associated ordinary differential equations (ODEs) governing multi-point kinetics formulation can be solved using numerical methods to the desired level of accuracy and thus allows formulation based on user defined control variables, i.e., without disturbing the source code and hence also avoiding associated coupling issues. Euler's method has been used in the present formulation to solve several coupled ODEs internally at each time step. The results have been verified against inbuilt point-kinetics models of RELAP5 and validated against 3D kinetics code TRIKIN. The model was used to identify the critical break in RIH of a typical large PHWR core. The neutronic asymmetry produced in the core due to the system induced transient was effectively handled by the multi-point kinetics model overcoming the limitation of in-built point kinetics model

RELAP 4/MOD 6 boiling water nodalization study

International Nuclear Information System (INIS)

Sonneck, G.; Pfau, H.

1985-09-01

The risk of nuclear steam supply systems is dominated by the core melt accidents. The first step to a realistic assessment of these sequences is the successful prediction of a loss of coolant event in a test loop. One of the codes for that is RELAP 4/MOD 6 and one of the important options in this code is the nodalization. The base of this work is the test LOCA No. 1 FIX II in Studsvik (Sweden) which also served as the OECD International Standard Problem 15. This report discusses the influence of different nodalizations, of different distributions of pressure, water and structural heat as well as of different bubble rise options, break flow coefficients, and heat transfer time steps. The most important result is that a simple RELAP 4/MOD6 model with less than 10 volumes is able to predict an experiment as LOCA No. 1 in FIX II successfully using only a fraction of the usual computing time. (Author)

RELAP5/MOD2 models and correlations

International Nuclear Information System (INIS)

Dimenna, R.A.; Larson, J.R.; Johnson, R.W.; Larson, T.K.; Miller, C.S.; Streit, J.E.; Hanson, R.G.; Kiser, D.M.

1988-08-01

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 geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included

Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities

International Nuclear Information System (INIS)

Mangal, Amit; Jain, Vikas; Nayak, A.K.

2011-01-01

In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

Energy Technology Data Exchange (ETDEWEB)

Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

2016-10-15

Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in