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Sample records for capability incorporating relap5

  1. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  2. 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)

  3. Nuclear Hybrid Energy System Modeling: RELAP5 Dynamic Coupling Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Nolan Anderson; Haihua Zhao; Shannon Bragg-Sitton; George Mesina

    2012-09-01

    The nuclear hybrid energy systems (NHES) research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation.

  4. RELAP5/MOD3 code coupling model

    International Nuclear Information System (INIS)

    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

  5. Evaluation of M-RELAP5 code capability for small-break LOCA analysis

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries, Ltd. (MHI) has developed M-RELAP5 code to analyze Small-Break LOCA of PWR for licensing calculations. MHI specifically selected RELAP5-3D and modified it as M-RELAP5 in order to meet the requirements in 10CFR Part 50 Appendix K, 'ECCS Evaluation Models'. MHI conducted several analyses for separate effect tests (i.e. ORNL/THTF, FLECHT, CCFL etc.) and for integral effect tests (i.e. ROSA/LSTF, LOFT and Semiscale) to investigate the applicability of M-RELAP5 for the important phenomena (CHF/core dryout, uncovered core heat transfer, core mixture level etc.) under the small-break LOCA, and for the validation against the system effects. The results show that M-RELAP5 well predicts the key phenomena and calculates core heatup conservatively. It is concluded that M-RELAP5 can be applied to the licensing calculation for small-break LOCA of PWR. (author)

  6. Incorporation of lithium lead eutectic as a working fluid in RELAP5 and preliminary safety assessment of LLCS

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K., E-mail: trivedi@iitk.ac.in [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India); Sandeep, K.T. [Institute for Plasma Research, Gandhinagar 382428 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India); Chaudhari, V.; Kumar, E.R. [Institute for Plasma Research, Gandhinagar 382428 (India); Munshi, P. [Nuclear Engineering and Technology Programme, Indian Institute of Technology, Kanpur 208016 (India)

    2014-12-15

    Highlights: • The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in ITER. • It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. • RELAP steady state results closely match with the operating conditions of LLCS. • Results from transient calculations show that a maximum temperature of 875 K is attained 300 s after the loss of LLE flow. - Abstract: The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE)

  7. RESTRUCTURING RELAP5-3D FOR NEXT GENERATION NUCLEAR PLANT ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; George L. Mesina; Joshua M. Hykes

    2006-06-01

    RELAP5-3D is used worldwide for analyzing nuclear reactors under both operational transients and postulated accident conditions. Development of the RELAP code series began in 1975 and since that time the code has been continuously improved, enhanced, verified and validated [1]. Since RELAP5-3D will continue to be the premier thermal hydraulics tool well into the future, it is necessary to modernize the code to accommodate the incorporation of additional capabilities to support the development of the next generation of nuclear reactors [2]. This paper discusses the reengineering of RELAP5-3D into structured code.

  8. RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa V. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg; Stefanova, Antoaneta E. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin P. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg

    2005-08-15

    During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

  9. RELAP5 capabilities in thermal-hydraulic prediction of SBWR containment behaviour: PANDA steady state and transient tests evaluation

    International Nuclear Information System (INIS)

    This paper summarizes the results of the qualification activity of RELAP5/Mod3.2 code performed using PANDA steady state and integral test experimental data. The steady state tests evaluate the PCC performances in removing decay heat power in presence and in absence of non-condensable gases, while the considered integral test (M3) simulates the transient following a break in the main steam line of the SBWR, using, as nominal initial conditions, those calculated for the SBWR under SSAR assumptions at one hour into the LOCA. The results obtained simulating both types of tests show a rather good and robust overall code behavior both in the simulation of steady state test and in the representation of the integral test considered: most of the main experimental results (WW/DW pressures, PCC heat exchange) were well represented by the code. The different studies performed indicated that: Different models of PCC pool lead a different trend of system pressure, and sometimes to an opening of vacuum breaker valves, that does not occur in the transient; The code underestimate the heat exchanged between PCC pool and tubes: n the considered test the system pressure is slightly overestimated (maximum 2% more than the experimental value). This fact is also proved by the differences in the temperature of the condensing mixture in the PCC, quite large in all the performed studies; The treatment of the non condensable gases, as implemented in the code, lead some errors in the calculation of the heat transfer coefficient in the PCC components and generally slow down the overall calculation. In general terms, the RELAP5/Mod3.2 was found to be suitable to represent the SBWR containment behavior under the conditions specified in the experimental side. (author)

  10. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable 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 RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  11. RELAP5/MOD2 code assessment

    International Nuclear Information System (INIS)

    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

  12. Development Program of LOCA Licensing Calculation Capability with RELAP5-3D in Accordance with Appendix K of 10 CFR 50.46

    International Nuclear Information System (INIS)

    In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during LOCA, it will take many more resources to develop and to get final approval from the licensing authority. Instead, implementation of evaluation models required by Appendix K of the Code of Federal Regulations, Title 10, Part 50 (10 CFR 50), upon an advanced thermal-hydraulic platform can also gain significant margin on the PCT calculation. A program to modify RELAP5-3D in accordance with Appendix K of 10 CFR 50 was launched by the Institute of Nuclear Energy Research, Taiwan, and it consists of six sequential phases of work. The compliance of the current RELAP5-3D with Appendix K of 10 CFR 50 has been evaluated, and it was found that there are 11 areas where the code modifications are required to satisfy the requirements set forth in Appendix K of 10 CFR 50. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effect experiments and six sets of integral-effect experiments will be adopted. Through the assessments program, all the model changes will be verified

  13. Streamlining of the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

    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

  14. RELAP5/MOD1 analysis of Turbine Trip Test 1 at Peach Bottom Unit 2

    International Nuclear Information System (INIS)

    The RELAP5/MOD1, Cycle 14 thermal-hydraulics code was used for analysis of turbine trip transient 1 (TT1) conducted at the Peach Bottom Atomic Power Station Unit 2. The goals were to evaluate the capability of RELAP5/MOD1 for BWR plant transient analysis and to evaluate the adaptability of RETRAN input models for use in RELAP5. The RELAP5/MOD1 input model used in the evaluation was derived from an existing RETRAN input model of Peach Bottom. Alteration of the RETRAN input model was required because of the differences between the codes. Modification of the RELAP5/MOD1, Cycle 14 code was also required to permit effective modeling of jet pumps and the steam separator. The results obtained with the RELAP5/MOD1, Cycle 14 code were in good agreement with the test data. Additional RELAP5 calculations were performed to determine the sensitivity of the results to variations in selected input parameters

  15. RELAP5 Assessment Using ISP38 Test Data

    International Nuclear Information System (INIS)

    The International Standard Problem (ISP) 38 experiment performed at the BETHSY integral test facility was calculated using two versions of the RELAP5 thermal-hydraulic code, RELAP5/MOD3.2.2gamma and RELAP5/MOD3.3beta. The experiment simulates a loss of residual heat removal system during mid-loop operation transient at 0.5 % (138 kW) of nominal core power value, with the pressurizer and steam generator outlet plenum manways open. The secondary side is full of air and isolated. Two nodalizations of the BETHSY facility were used to perform the calculations with both versions of the code, the first was developed for the RELAP5/MOD2 code and modified for the newer versions, the second was developed for the RELAP5/MOD3.2.2 code using the 'sliced' approach and is more detailed. The results of the calculations were compared with the experimental data and conclusions were drawn about the effect of the more detailed nodalization on the calculation results for both code versions. Several sensitivity studies were performed to evaluate the effects of changing the selected parameters on the calculated results. The analysis of the obtained results was used to give an evaluation of the capability of the RELAP5 code to simulate mid-loop operation transients and its relevance for the safety analysis of NPP Krsko. (author)

  16. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    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.

  17. Architectural Advancements in RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L. Mesina

    2005-11-01

    As both the computer industry and field of nuclear science and engineering move forward, there is a need to improve the computing tools used in the nuclear industry to keep pace with these changes. By increasing the capability of the codes, the growing modeling needs of nuclear plant analysis will be met and advantage can be taken of more powerful computer languages and architecture. In the past eighteen months, improvements have been made to RELAP5-3D [1] for these reasons. These architectural advances include code restructuring, conversion to Fortran 90, high performance computing upgrades, and rewriting of the RELAP5 Graphical User Interface (RGUI) [2] and XMGR5 [3] in Java. These architectural changes will extend the lifetime of RELAP5-3D, reduce the costs for development and maintenance, and improve it speed and reliability.

  18. A three-dimensional nodal neutron kinetics capability for relaps

    International Nuclear Information System (INIS)

    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

  19. Presentation of RELAP5 results on the personal computer

    International Nuclear Information System (INIS)

    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)

  20. BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code

    International Nuclear Information System (INIS)

    The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, 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 overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)

  1. RELAP5/MOD2 overview and developmental assessment results from TMI-1 plant transient analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

    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. 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 subcooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2.

  2. A study on CANDU model assessment of RELAP5/CANDU using RD-14M B9401 multi-channel RIH break experiment

    International Nuclear Information System (INIS)

    B9401 experiment, performed in RD-14M[1] multi-channel facility, was analyzed using RELAP5/MOD3 and RELAP5/CANDU and compared with experiment results. The RELAP5/CANDU code has been developed since 1998, based on RELAP5, in order to have auditing tool of CANDU NPP. The RELAP5/CANDU code is under developing and they have not been assessed much for a CANDU reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14M facility and to get the assessment results for RELAP5/CANDU code. The RD-14M test facility at Whiteshell Nuclear Research Establishment is a full-scale multi-channel pressurized-water loop. The RELAP5/MOD3 and RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the main phenomena occurred during the transient, in qualitative view. In quantitative view, the RELAP5/CANDU[4] predicted better than that of RELAP5. In the case of experiment that the stratification in fuel channel is dominant, it is expected that RELAP5/CANDU can give more accurate result than RELAP5

  3. Recent SCDAP/RELAP5 code applications and improvements

    International Nuclear Information System (INIS)

    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

  4. Recent SCDAP/RELAP5 code applications and improvements

    Energy Technology Data Exchange (ETDEWEB)

    Harvego, E.A.; Ghan, L.S.; Knudson, D.L.; Siefken, L.J. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

    1998-03-01

    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.

  5. PACTEL ISP-33. RELAP5 assessment

    International Nuclear Information System (INIS)

    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.)

  6. 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)

  7. Data calculation program for RELAP 5 code

    International Nuclear Information System (INIS)

    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)

  8. 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.

  9. RELAP5 SCDAP RIA transients analysis

    International Nuclear Information System (INIS)

    Some Romanian research programs require the study of fuel behavior during transients.. Romanian TRIGA Annular Core Pulse Reactor (ACPR) is provided with capabilities to produce short pulses that simulate Reactivity Initiated Accidents (RIA). A program was initiated to study CANDU type fuel in transient situations by using these reactor capabilities. Since 1984, over 54 tests were performed with fresh un-irradiated fuel specimens, placed in a special capsule. The goal of these tests was to determine the clad failure threshold and to determine fuel-clad-coolant interaction during transients. The irradiation tests conditions for this special capsule are atmospheric pressure and environmental temperature. This capsule is placed in the ACPR central dry cavity during the tests. During the pulse, in a very short time a significant energy is deposed in the fuel, which heats up the fuel. The clad temperature rises to over 1000 C. Due to the high temperature; the clad will quickly be covered with a vapor layer. Subsequent cooling of the clad is described as rewetting, involving thermal hydraulic phenomena which take place in the re-flooding phase following a LOCA accident, when on the fuel elements, covered with steam, cold water is injected from the emergency cooling system. The capsule and the fuel specimen was modeled using RELAP5-SCDAP code. To document our SCDAP model we used some JAERI experiments. These evaluations show a good agreement. The SCDAP model developed for clad temperature prediction for the tests in Romania's ACPR, with the atmospheric capsule and fresh fuel, as well as for similar JAERI reactor (Japan) tests proved to have good predictions and, also, gave a good prediction for clad failure threshold energy. The predicted failure threshold is confirmed both for the ACPR tests and for the JAERI tests

  10. SCDAP/RELAP5 independent peer review

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 ftb' 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

  12. A comparative assessment result of B9401 multi-channel RIH break experiment with Canadian test facility RD-14M using RELAP5/MOD3 and RELAP5/CANDU+

    International Nuclear Information System (INIS)

    The experiment, B9401, performed in RD-14M multi-channel experimental facility, was preliminary analyzed using RELAP5/MOD3 and RELAP5/CANDU+ and compared with experimental results. The RELAP5 code has been developed for best-estimate transient simulation of pressurized water reactors and associated system, but the RELAP5/CANDU+ code has been developed since 1998 in order to have auditing tool of CANDU NPP. The RELAP5/CANDU+ code is under developing and they have not been assessed much for a CANDU reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14M facility and to get the assessment results for RELAP5/CANDU+ code. The RD-14M test facility at Whiteshell Nuclear Research Establishment is a full-scale multi-channel pressurized-water loop. The RD-14M is not a 'scale' model of any particular CANDU reactor. It possesses many geometric features of a CANDU reactor heat transport system, and is capable of operating at conditions similar to those expected to occur in a reactor under normal operation and some postulated accident conditions. As preliminary results, the RELAP5/MOD3 and RELAP5/MOD+ analyses demonstrate the code's capability to predict reasonably the main phenomena occuring in the transient, in qualitative view. In quantative view, the RELAP5/CANDU+ predicted better than that of RELAP5. However, some discrepancies after emergency coolant injection, the behaviors of the ECI mass flow rate and the sheath temperatures were observed commonly

  13. Validation of RELAP5 critical flow model

    International Nuclear Information System (INIS)

    The critical two-phase flow computerized simulation is made with the RELAP5 computer code. The methodology breaks the chocking process into one with either a two-phase inlet or a subcooled inlet. For two-phase flow the phases are assumed to be in thermal equilibrium. Thermal non-equilibrium is considered for subcooled upstream stagnation conditions. (authors); 6 refs., 3 figs

  14. Coupled RELAP5/GOTHIC model for IRIS SBLOCA analysis

    International Nuclear Information System (INIS)

    . 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)

  15. RELAP5-3D Restart and Backup Verification Testing

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2013-09-01

    Existing testing methodology for RELAP5-3D employs a set of test cases collected over two decades to test a variety of code features and run on a Linux or Windows platform. However, this set has numerous deficiencies in terms of code coverage, detail of comparison, running time, and testing fidelity of RELAP5-3D restart and backup capabilities. The test suite covers less than three quarters of the lines of code in the relap directory and just over half those in the environmental library. Even in terms of code features, many are not covered. Moreover, the test set runs many problems long past the point necessary to test the relevant features. It requires standard problems to run to completion. This is unnecessary for features can be tested in a short-running problem. For example, many trips and controls can be tested in the first few time steps, as can a number of fluid flow options. The testing system is also inaccurate. For the past decade, the diffem script has been the primary tool for checking that printouts from two different RELAP5-3D executables agree. This tool compares two output files to verify that all characters are the same except for those relating to date, time and a few other excluded items. The variable values printed on the output file are accurate to no more than eight decimal places. Therefore, calculations with errors in decimal places beyond those printed remain undetected. Finally, fidelity of restart is not tested except in the PVM sub-suite and backup is not specifically tested at all. When a restart is made from any midway point of the base-case transient, the restart must produce the same values. When a backup condition occurs, the code repeats advancements with the same time step. A perfect backup can be tested by forcing RELAP5 to perform a backup by falsely setting a backup condition flag at a user-specified-time. Comparison of the calculations of that run and those produced by the same input w/o the spurious condition should be

  16. Sub-channel analysis by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Alessandro Petruzzi; Anis Bousbia Salah [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2005-07-01

    Full text of publication follows: Recent progress in computer technology has increased the possibilities for code calculations in predicting realistically transient scenarios in nuclear power plants. Several attempts have been engaged in order to enlarge the domain for code applications, and to allow best estimate core simulation including interaction effects between neutronics and thermal-hydraulics. In this context, Relap5/Mod3.3 system thermalhydraulic code was used as a sub-channel code for the simulation of the low-pressure boil off experiment No 5002 of Neptun test facility. The experiment constitutes one of the separate effects test (SET) in the OECD/CSNI matrix for thermalhydraulic code validation related to phase separation and vertical flow 'with or without mixture level'. The drying out of the heated elements is expect to occur at very low coolant flow rates, low pressure (about 1.1 bar) and low power level (24.6 kW). The main aim of the activity discussed in the paper is to develop a 'nodalization technology' for accurately modeling the sub-channel grade void distribution problem and in the same way to assess the degree of success in using the Relap5 system code as a sub-channel code for the analysis of local quantities during transients in nuclear reactors. All thermal-hydraulic parameters, such as the collapsed liquid level, critical heat flux time occurrence and heaters surface temperature have been predicted with reasonable accuracy. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. More accurate results have been obtained considering the surface to surface radiation heat transfer model, as well as more cross flow nodes between the test section rods. The overall analysis confirms the possibility of using the Relap5/Mod3.3 system thermal-hydraulic code as sub-channel code to predict the evolution of relevant local quantities measured during 'relevant' experiments

  17. AUTOMATED, HIGHLY ACCURATE VERIFICATION OF RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    George L Mesina; David Aumiller; Francis Buschman

    2014-07-01

    Computer programs that analyze light water reactor safety solve complex systems of governing, closure and special process equations to model the underlying physics. In addition, these programs incorporate many other features and are quite large. RELAP5-3D[1] has over 300,000 lines of coding for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. Verification ensures that a program is built right by checking that it meets its design specifications. Recently, there has been an increased importance on the development of automated verification processes that compare coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions[2]. For the first time, the ability exists to ensure that the data transfer operations associated with timestep advancement/repeating and writing/reading a solution to a file have no unintended consequences. To ensure that the code performs as intended over its extensive list of applications, an automated and highly accurate verification method has been modified and applied to RELAP5-3D. Furthermore, mathematical analysis of the adequacy of the checks used in the comparisons is provided.

  18. RELAP5-3D Code Includes ATHENA Features and Models

    International Nuclear Information System (INIS)

    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, SF6, 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)

  19. Analysis of ROSA-III test RUN 704 by RELAP5/MOD0 code

    International Nuclear Information System (INIS)

    The ROSA-III test RUN 704 was analyzed for the assessment of RELAP5 code for BWR LOCA. RELAP5 is an advenced code developed to analyze thermal-hydraulic phenomena during LOCA and non-LOCA transients of LWR. It is based on a one-dimensional, nonhomogeneous, nonequilibrium two-phase flow model. The ROSA-III test RUN 704 is a standard BWR LOCA test, simulating a 200% double-ended break at the recirculation pump inlet pipe with all emergency core cooling systems activated. Large increase of core inlet flow due to lower plenum flashing and resulted rewetting of heater surface were calculated by RELAP5, indicating superior capability of RELAP5 two-phase flow model than RELAP4 phase separation model. Vapor and liquid counter-current flow was calculated at core inlet and core outlet. A small degree thermal nonequilibrium between vapor and liquid was calculated in the upper plenum after HPCS activation. However, core reflooding and quenching of heater surface were not calculated. There are still room for improvement in the interfacial drag and the heat transfer models of RELAP5/MOD0. (author)

  20. Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2006-01-01

    Full Text Available RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

  1. RELAP5-3D Code Includes Athena Features and Models

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

    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, sf6, 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.

  2. Preliminary design report for SCDAP/RELAP5 lower core plate model

    International Nuclear Information System (INIS)

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  3. Preliminary design report for SCDAP/RELAP5 lower core plate model

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Griffin, F.P. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

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

    International Nuclear Information System (INIS)

    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)

  5. Analysis of the reflood experiment by RELAP5/MOD2 code

    International Nuclear Information System (INIS)

    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)

  6. RELAP5 analysis of two-phase decompression and rarefaction wave propagation under a temperature gradient

    International Nuclear Information System (INIS)

    The capability of RELAP5 to model single and two-phase acoustic waves is demonstrated with the use of fine temporal and spatial discretizations. Two cases were considered: a single phase air shock tube problem and pressure waves observed by Takeda and Toda in a two-phase decompression experiment in a pipe. Whereas the agreement for the single phase case is excellent, some discrepancies were observed in the two-phase case. However, RELAP5 produced markedly better results after adjusting the bubble size and the choked flow area. These results illustrate the need of a dynamic model for the interfacial area concentration (i.e., the bubble size). (author)

  7. Preliminary study of coupling CFD code FLUENT and system code RELAP5

    International Nuclear Information System (INIS)

    Highlights: • System code RELAP5/MOD3.1 is coupled with CFD code FLUENT through DLL and UDF. • Transient water flow in a simple straight tube is tested using the coupled tool. • Simulation of Edwards’ pipe blowdown experiment using the coupled tool is conducted. • Coupled analysis of a more comprehensive thermal–hydraulic system is performed. - Abstract: The present paper discusses a coupling strategy of the 3D (three-dimensional) computational fluid dynamics (CFD) code ANSYS-FLUENT with the best estimate 1D (one-dimensional) thermal–hydraulic system code RELAP5/MOD3.1. Preliminarily, by using DLL (Dynamic Link Library) technology and FLUENT UDF (User Defined Functions), an explicit coupling method expected to be able to support the analysis of multi-purpose thermal–hydraulic phenomena in nuclear reactor systems has been developed. Calculations for two test cases using the coupled FLUENT/RELAP5 code have been carried out to test and demonstrate the coupling capability: (i) the first one consisting of single-phase water transient flow in a square straight tube with well controlled mass flow rates; (ii) the second one illustrating the process of single-phase water flow in a system including two closed loops and one vessel, on which loss of loop water flow due to pump trip and increase of loop water temperature are studied. Both reasonable 1D systematic behaviors and 3D distribution information are naturally obtained for the test cases. Besides, a study of a highly transient experiment problem, i.e. Edwards–O’Brien pipe blowdown problem, has been performed by using the coupled FLUENT/RELAP5 code. The results are compared with standalone RELAP5 calculation and available experimental data, which shows the coupled FLUENT/RELAP5 code’s acceptable potential for the capability of analyzing either simple single-phase or complex two-phase flow problem

  8. Modification and validation of RELAP5/MOD3.2 for thermal-hydraulic accident analyses of HANARO

    International Nuclear Information System (INIS)

    Many aspects of RELAP5/MOD3.2 are modified by new features to properly simulate the HANARO characteristics such as the finned fuel elements and cooling by the plate type heat exchanger. Especially, RELAP5/MOD3.2 is modified a new heat transfer correlation package appropriate for the accident analysis of HANARO and developed plate type heat exchanger model. The validation of this newly developed code, RELAP5/HANARO, is carried out to assess its calculational capability to predict thermal-hydraulic behavior through the accident analysis in HANARO. The simulations performed are the single pin heat transfer and plate type heat exchanger experiments, which are then compared with experimental results and manufacturer's data, respectively. The simulation of natural convection experiment with the scaled bundle is also performed to evaluate the natural circulation cooling capability. The assessment results for the single pin heat transfer experiment and plate type heat exchanger model with RELAP5/HANARO showed conservatively good agreement with the experimental results. The natural circulation results calculated by RELAP5/HANARO are similar to the experimental data, even though minor discrepancies of the flow are identified. However, these differences are insignificant and conservatively acceptable thermal hydraulics. RELAP5/HANARO comprises the unique characteristics of HANARO and it is capable of simulating well the thermal-hydraulic behavior. In conclusion, it can be stated that this modified code provides a suitable analytical tool for the thermal-hydraulic accident analyses of HANARO

  9. Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations

    International Nuclear Information System (INIS)

    Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)

  10. RELAP5/MOD2 models and correlations

    International Nuclear Information System (INIS)

    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

  11. RELAP5/MOD2 models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Validation of the thermal hydraulic computer code S-RELAP5 for performing loss-of-coolant accident analysis (LOCA) in Pressurized Water Reactors (PWRs)

    International Nuclear Information System (INIS)

    Siemens Power Corporation (SPC) has developed S-RELAP5, a RELAP5/MOD2 based thermal hydraulic system code with main modifications and improvements relative to RELAP5/MOD2 concerning Multi-Dimensional Capability, Energy Equations, Numerical Solution of Hydrodynamic, Constitutive Models, Heat Transfer Models, Chocked Flow, and Counter-Current Flow Limiting. S-RELAP5 was exercised over a range of integral and separate effects tests in order to demonstrate that the code could predict the important phenomena associated with PWR LBLOCA. A methodology for calculation of statistical uncertainties has been developed and applied to analyses of hypothetical large break loss-of-coolant accidents (LBLOCA). To extend the application capability of S-RELAP5 to small break loss-of-coolant accidents problems (SBLOCA) an investigation program for appropriate experiments was launched and largely carried out. (author)

  13. Application of RELAP5 to a pipe blowdown experiment

    International Nuclear Information System (INIS)

    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

  14. Independent review of SCDAP/RELAP5 natural circulation calculations

    International Nuclear Information System (INIS)

    A review and assessment of the uncertainties in the calculated response of reactor coolant system natural circulation using the SCDAP/RELAP5 computer code were completed. The SCDAP/RELAP5 calculation modeled a station blackout transient in the Surry nuclear power plant and concluded that primary system depressurization from natural circulation induced primary system failure is more likely than previously thought

  15. Independent review of SCDAP/RELAP5 natural circulation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, G.M.; Gross, R.J.; Martinez, M.J.; Rightley, G.S.

    1994-01-01

    A review and assessment of the uncertainties in the calculated response of reactor coolant system natural circulation using the SCDAP/RELAP5 computer code were completed. The SCDAP/RELAP5 calculation modeled a station blackout transient in the Surry nuclear power plant and concluded that primary system depressurization from natural circulation induced primary system failure is more likely than previously thought.

  16. Simulation of water hammer experiments using RELAP5 code

    International Nuclear Information System (INIS)

    The rapid closing or opening of a valve causes pressure transients in pipelines. The fast deceleration of the liquid results in high pressure surges upstream the valve, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increases. This phenomenon is called water hammer. The intensity of water hammer effects will depend upon the rate of change in the velocity or momentum. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the thermal-hydraulic system since, if the pressure induced exceeds the pressure range of a pipe given by the manufacturer, it can lead to the failure of the pipeline integrity. Due to its potential for damage of pipes, water hammer has been a subject of study since the middle of the nineteenth century. Many theoretical and experimental investigations were performed. The experimental investigation of the water hammer tests performed at Fraunhofer Institute for Environmental, Safety and Energy Technology (UMSICHT) [1] and Cold Water Hammer experiment performed by Forschungszentrum Rossendorf (CWHTF) [2] should be mentioned. The UMSICHT facility in Oberhausen was modified in order to simulate a piping system and associated supports that are typical for a nuclear power plant [3]. The Cold water hammer experiment is interesting and instructive because it covers a wide spectrum of particularities. One of them is sub-cooled water interaction with condensing steam at the closed end of the vertical pipe at room temperature and corresponding saturation pressure [4]. In the paper, the capabilities of RELAP5 code to correctly represent the water hammer phenomenon are presented. Paper presents the comparison of RELAP5 calculated and measured at UMSICHT and CWHTF test facilities pressure transient values after the fast closure (opening) of valves. The analyses of rarefaction wave travels inside the pipe and condensation of vapour bubbles in the liquid column

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

    International Nuclear Information System (INIS)

    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

  18. MELCOR and SCDAP/RELAP5 code validation by simulation of TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Chris [400 Central Drive, School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Liao, Yehong; Vierow, Karen

    2005-07-01

    Full text of publication follows: A comparison of the two severe accident codes, MELCOR and SCDAP/RELAP5, within the scope of thermal-hydraulic and core degradation models, has been previously performed by the authors for a hypothetical station blackout severe accident of a typical 4-loop PWR. This paper describes a validation of the codes. In the simulation of the TMI-2 severe accident, the data recorded during the accident and inferred from post-accident phenomena were used to investigate the soundness and compare the capabilities of MELCOR and SCDAP/RELAP5. With versatile control functions, best estimate thermal-hydraulic component models and detailed core models, MELCOR and SCDAP/RELAP5 produced similar predictions of the progression of TMI-2 accident, in simulating the actions of the plant control room operators, the system thermal-hydraulic response, the fuel damages, the core degradation and relocation. Input models and assumptions were modified to be as consistent yet true to the actual plant as possible. Due to different development approaches and some unique models, some minor discrepancies were observed between the predictions of MELCOR and SCDAP/RELAP5, which are within the uncertainties of the code numerical computation and the physics models. Some significant discrepancies in a few key areas were resolved either with a sophisticated method comparing with phenomena instead of raw code output information, or with a model modification based on actual plant data and experiment results. (authors)

  19. RELAP5-3D Compressor Model

    International Nuclear Information System (INIS)

    A compressor model has been implemented in the RELAP5-3D code. The model is similar to that of the existing pump model, and performs the same function on a gas as the pump performs on a single-phase or two-phase fluid. The compressor component consists of an inlet junction and a control volume, and optionally, an outlet junction. This feature permits cascading compressor components in series. The equations describing the physics of the compressor are derived from first principles. These equations are used to obtain the head, the torque, and the energy dissipation. Compressor performance is specified using a map, specific to the design of the machine, in terms of the ratio of outlet-to-inlet total (or stagnation) pressure and adiabatic efficiency as functions of rotational velocity and flow rate. The input quantities are specified in terms of dimensionless variables, which are corrected to stagnation density and stagnation sound speed. A small correction was formulated for the input of efficiency to account for the error introduced by assumption of constant density when integrating the momentum equation. Comparison of the results of steady-state operation of the compressor model to those of the MIT design calculation showed excellent agreement for both pressure ratio and power

  20. Development of the unified version of COBRA/RELAP5

    International Nuclear Information System (INIS)

    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

  1. 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)

  2. Mathematic preprocessor for RELAP5 code using Microsoft Excel

    International Nuclear Information System (INIS)

    Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a toll for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-Excel software. (author)

  3. Transient simulation of ALWR passive safety systems using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

  4. A containment convective loop analysis using the RELAP5-Mod 3.2

    International Nuclear Information System (INIS)

    The present study was performed to verify the RELAP5-Mod 3.2 code capability to calculate convection phenomena of the type occurring in a convective loop. A simplified geometrical model of a reactor containment system was used. The parametric studies were made for the main variables which govern material transport in the volume junctions considered. The results obtained and that got using the same model with the CONTAIN code, were compared. The comparison is satisfactory. (author). 3 refs., 11 figs

  5. Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes

    Directory of Open Access Journals (Sweden)

    Viet-Anh Phung

    2015-01-01

    Full Text Available In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified.

  6. Pump-stopping water hammer simulation based on RELAP5

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    International Nuclear Information System (INIS)

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  9. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Jung, B. D.; Lee, W. J.; Lim, H. S. [KAERI, Taejon (Korea, Republic of); Bang, Y. S.; Kim, M. W.; Lee, S. H. [KINS, Taejon (Korea, Republic of)

    1999-10-01

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations.

  10. Improvement of RELAP5/MOD3.2.2 models for the development of CANDU auditing code

    International Nuclear Information System (INIS)

    Thermal-hydraulic models of the model improvements of NRC PWR auditing tool , i.e. RELAP5/MOD3, current auditing tool for LWR licensing, have been improved were attempted to develop a than auditingermal hydraulic auditing code for the CANDU. Major in order to identify the thermal hydraulic phenomena for the key CANDU events for key eventswere identified for reactor systems and components. Based on this, the modeling limitation of current RELAP5/MOD3 for CANDU applications were derived and the model improvement areas were identified. By improving these models in the code, RELAP5/MOD3/CANDU version has been developed. The new version was written in FORTRAN90 and its application capability has been verified through the simple verification calculations

  11. Development and assessment of the COBRA/RELAP5 code

    International Nuclear Information System (INIS)

    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

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@mail.polimi.i [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Grgic, Davor [Department of Power Systems, FER, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Cammi, Antonio; Ricotti, Marco E. [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy)

    2011-04-15

    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.

  14. Assessment study of RELAP5/MOD2 Cycle 36.04 based on the DOEL-4 manual loss of load test of November 23, 1985

    International Nuclear Information System (INIS)

    The loss of external load test conducted on the DOEL-4 power plant has been analyzed on the basis of a high quality data acquisition system. A detailed numerical analysis of the transient by means of the best estimate code RELAP5/MOD2 is presented. The RELAP5 code is capable to simulate the basic plant behavior. Deficiencies noted involved structural heat simulation, acoustic phenomena, and excessive interphase drag

  15. SCDAP/RELAP5 code development and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Hohorst, J.K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1996-03-01

    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.

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

    International Nuclear Information System (INIS)

    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)

  17. Study of turbine simulation model based on RELAP5

    International Nuclear Information System (INIS)

    The turbine model which can represent accurately non-isentropic process in the stage of turbine and system dynamic characteristics was developed and added into RELAP5 code, and the improvement of the turbine model of RELAP5 was implemented. The improved turbine model is based on the characteristics of steam flow and work transfer in the stage of turbine and considers adequately the impact of internal configuration parameters and oblique shock which is developed by non-equilibrium condensation of wet steam in turbine. Through building internal coupling interface and the modifying input processing subroutines, turbine model was developed as a part of RELAP5 hydro dynamic model. Taking the turbine of Qinshan 300 MW Nuclear Power Plant as an example, the simulation calculation and comparative analysis were performed for both stead and dynamic cases respectively by both the original and the modified turbine models in RELAP5 code. The results show that the modified turbine model can represent more accurately the dynamic operation characteristics of the turbine. (authors)

  18. RELAP5/MOD3 simulation of the water cannon phenomenon

    International Nuclear Information System (INIS)

    The results of the transient behavior of the water cannon phenomenon determined by RELAP5/MOD3 Version 5m5 are presented. The physical system consists of a 0.7112-m-long, 0.381-m-i.d. vertical tube partially immersed in a reservoir of subcooled water. The tube is closed at the top and initially filled with saturated steam. The water cannon is created when a liquid slug is drawn in to the tube because of the rapid condensation of the steam. In a fraction of a second, the liquid slug strikes the top end of the tube and causes a large pressure spike. The primary objective is to apply the RELAP5/MOD3 computer code to analyze the water cannon event and assess the ability of RELAP5/MOD3 to simulate fast two-phase transients. The sensitivity of time-step size and mesh size has been studied. It is found that RELAP5/MOD3 adequately simulated the transient process with a mesh size of 0.07112 m (i.e., ten nodes) and a time-step size of 10-5 s. The calculated peak pressure of the first pressure spike is of the same order of magnitude as experimental data from literature. The effect of reservoir temperature on the magnitude of the first pressure spike is also studied, and it is found that the pressure peak value decreased with increasing reservoir temperature

  19. RELAP5-3D Developer Guidelines and Programming Practices

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2014-03-01

    Our ultimate goal is to create and maintain RELAP5-3D as the best software tool available to analyze nuclear power plants. This begins with writing excellent programming and requires thorough testing. This document covers development of RELAP5-3D software, the behavior of the RELAP5-3D program that must be maintained, and code testing. RELAP5-3D must perform in a manner consistent with previous code versions with backward compatibility for the sake of the users. Thus file operations, code termination, input and output must remain consistent in form and content while adding appropriate new files, input and output as new features are developed. As computer hardware, operating systems, and other software change, RELAP5-3D must adapt and maintain performance. The code must be thoroughly tested to ensure that it continues to perform robustly on the supported platforms. The coding must be written in a consistent manner that makes the program easy to read to reduce the time and cost of development, maintenance and error resolution. The programming guidelines presented her are intended to institutionalize a consistent way of writing FORTRAN code for the RELAP5-3D computer program that will minimize errors and rework. A common format and organization of program units creates a unifying look and feel to the code. This in turn increases readability and reduces time required for maintenance, development and debugging. It also aids new programmers in reading and understanding the program. Therefore, when undertaking development of the RELAP5-3D computer program, the programmer must write computer code that follows these guidelines. This set of programming guidelines creates a framework of good programming practices, such as initialization, structured programming, and vector-friendly coding. It sets out formatting rules for lines of code, such as indentation, capitalization, spacing, etc. It creates limits on program units, such as subprograms, functions, and modules. It

  20. RELAP5 analysis of PKL, main steam line break test

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bencik, V.; Debrecin, N.; Grgic, D. [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Bajs, T. [Enconet International Ltd, Miramarska 20, 10000 Zagreb (Croatia)

    2010-07-01

    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)

  3. More conservative governing equations in RELAP5: Derivation of equations

    International Nuclear Information System (INIS)

    Highlights: • The “non-conservative” numerical approximation is used in current versions of RELAP5. • Mass and energy errors increase for some transients due to non-conservativism. • This paper shows the derivation of a numerical approach to eliminate the mass error. • The second article (Fu et al., 2015) shows the (strategic) solution for the approach. - Abstract: The design and analysis of the thermal/hydraulic systems in nuclear power plants necessitate system codes that can be used in the analysis of steady state and transient conditions. RELAP5 is one of the most commonly used system codes in nuclear organizations. RELAP5 is based on a two-fluid, non-equilibrium, non-homogeneous, hydrodynamic model for the transient simulation of the two-phase system behavior. This model includes six governing equations to describe the mass, energy, and momentum of the two fluids. The “non-conservative” numerical approximation form (which is the current version of RELAP5 code) is obtained through the manipulation of selected derivative terms in the equations including the linearization of the product terms in the time derivatives of the equations. In the non-conservative technique, the truncation errors introduced in the linearization process can produce mass and energy errors for some classes of transients during time advancements, either resulting in (a) automatic reduction of time steps used in the advancement of the equations and increased run times or (b) the growth of unacceptably large errors in the transient results. To eliminate these difficulties, an optional numerical approach has been introduced in RELAP/SCDAPSIM/MOD4.0. This approach uses a more consistent set of “conservative” numerical approximations to solve non-linearized mass and energy governing equations. The RELAP/SCDAPSIM/MOD4.0 code, being developed as part of the international Severe Core Damage Analysis Package (SCDAP) Development and Training Program (SDTP), is the first version of

  4. Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. A demonstration calculation is presented. Finally, future tasks and plans are outlined. (author)

  5. Coupling the RELAP5-3d advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. Finally, future tasks and plans are outlined. (author)

  6. Qualification of the code RELAP5/MOD2 with regard to pressurizer separate effect experiments

    International Nuclear Information System (INIS)

    As part of the RELAP5/MOD2 code assessment matrix, developed to qualify this simulation tool for analysis of pressurization transients in DWRs, two pressurizer separate effect experiments were analysed. The results allow some conclusions concerning simulation of pressurization transients in real nuclear plants. The geometry, thermal properties and heat losses of the pressurizer as well as the time lag constant of the instrumentation and the time step of the calculation were identified as the key parameters. Some conclusions were obtained concerning the code's capability to predict the thermal gradients, the heat transfer at the different interfaces, the condensation and evaporation rates, and their impact on pressure behaviour. (orig.)

  7. Qualification of the code RELAP5/MOD2 with regard to pressurizer separate effect experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1992-12-01

    As part of the RELAP5/MOD2 code assessment matrix, developed to qualify this simulation tool for analysis of pressurization transients in DWRs, two pressurizer separate effect experiments were analysed. The results allow some conclusions concerning simulation of pressurization transients in real nuclear plants. The geometry, thermal properties and heat losses of the pressurizer as well as the time lag constant of the instrumentation and the time step of the calculation were identified as the key parameters. Some conclusions were obtained concerning the code's capability to predict the thermal gradients, the heat transfer at the different interfaces, the condensation and evaporation rates, and their impact on pressure behaviour. (orig.).

  8. Evaluation of a Jet Pump Model for RELAP5 code

    International Nuclear Information System (INIS)

    The report presents the results of revision and evaluation of the JAERI Jet Pump Model which has been developed for RELAP5 code. Analyses for the ROSA-III experiments were performed by the RELAP5 code with the Jet Pump Model. The model can solve not only the single phase steady water flows but also the transient flows through the ROSA-III jet pumps. The JAERI Jet Pump Model can predict the jet pump characteristics more accurately than the small pump model used for convenience to drive the suction flow. And therefore, the JAERI Jet Pump Model is useful to calculate accurately the mass inventory in each region and heat transfer in core. The model for the BWR jet pump analyses, however, needs appropriate geometrical data which represent the complicated suction flow paths around the 5 holes drive nozzle. (author)

  9. Peer review of RELAP5/MOD3 documentation

    International Nuclear Information System (INIS)

    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

  10. Integral experiment and RELAP5 analysis for DVI line break SBLOCA in APR1400

    International Nuclear Information System (INIS)

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in the pressurized water reactor, APR1400, were investigated. To understand the thermal-hydraulic phenomena during the SBLOCA transient, the reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed according to the energy scaling methodology. The methodology conserves the mass inventory and energy of the system in the same time scale as the prototype. From the RELAP5 analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In order to overcome the limitation of power in actual SNUF, the modified-power curve was utilized without simulating the forced flow by pump, so that those corrections did not affect the major phenomena during transient. Geometric distortion of actual SNUF also did not strongly disturb the thermal-hydraulic behaviors, especially occurrence of the downcomer seal clearing. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. It occurred when the steam injected from cold legs penetrated the coolant in upper downcomer toward the broken DVI line. The experimental results was used to validate the calculation capability of RELAP5, especially for the downcomer seal clearing phenomenon, and to estimate the scale-up capability of RELAP5 code according to the scaling methodology. (authors)

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

    International Nuclear Information System (INIS)

    feature is available for PC Windows users and provides simple Graphic User Interface (GUI) features. The productivity gains for both new, and experienced users from this userfriendly interface will be enormous, and the increased user productivity will pay back the developmental costs. RELAP5/MOD3.2.1.2 has been moderized and restructured in order to enhance the code portability, maintenance capability, readability, and flexibility. User convenience for PC Windows users has been realized by the on-line graphical processing through Windows programming. It should be noted that the code strcuture was fully domesticated, and future improvements could be easily carried out with the restructured version of RELAP5/MOD3.2.1.2

  12. RELAP5-3D Resolution of Known Restart/Backup Issues

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George L.; Anderson, Nolan A.

    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.

  13. Web-based, Interactive, Nuclear Reactor Transient Analyzer using LabVIEW and RELAP5 (ATHENA)

    International Nuclear Information System (INIS)

    In nuclear engineering, large system analysis codes such as RELAP5, TRAC-M, etc. play an important role in evaluating a reactor system behavior during a wide range of transient conditions. One limitation that restricts their use on a wider scale is that these codes often have a complicated I/O structure. This has motivated the development of GUI tools for best estimate codes, such as SNAP and ViSA, etc. In addition to a user interface, a greater degree of freedom in simulation and analyses of nuclear transient phenomena can be achieved if computer codes and their outputs are accessible from anywhere through the web. Such a web-based interactive interface can be very useful for geographically distributed groups when there is a need to share real-time data. Using mostly off-the-shelf technology, such a capability - a web-based transient analyzer based on a best-estimate code - has been developed. Specifically, the widely used best-estimate code RELAP5 is linked with a graphical interface. Moreover, a capability to web-cast is also available. This has been achieved by using the LabVIEW virtual instruments (VIs). In addition to the graphical display of the results, interactive control functions have also been added that allow operator's actions as well as, if permitted, by a distant user through the web

  14. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    OpenAIRE

    Bousbia-Salah Anis; D'auria Francesko

    2002-01-01

    This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to...

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

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. Loss Of Secondary coolant accident analysis for Pius type reactor using Relap5/MOD2

    International Nuclear Information System (INIS)

    Process inherent ultimate safety (Pius) reactor concept is a reactor concept that intrinsically based on passive safety. This reactor refer to Pressurized water reactor (PWR) wherein the primary system is submerged in a pool of poison water. the operating principle is to maintain the pressure balance, so that no inflow from pool to the primary system. On loss of secondary coolant accident, primary coolant temperature increases, it is followed by the increase of primary pump speed. When the upper limit is reached, the pump is tripped due to the pressure balance disturbance, poison water flows from pool to the primary system, then reactor shut down. this accident condition was simulated by experimental and numerical simulation using RELAP5/MOD2. Numerical simulation was done to the experimental apparatus nodalization that was set on the norm of RELAP5/MOD2. This nodalization consist of 119 volumes, 127 junctions, and 106 heat structures. Analysis was carried out using both experimental and numerical simulation results. it can be concluded that PIUS type reactor is able to anticipate loss of secondary coolant accident because its capability of self shut down

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

    International Nuclear Information System (INIS)

    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

  19. Validation of RELAP5 with sensitivity analysis for uncertainty assessment for natural circulation two-phase flow instability

    International Nuclear Information System (INIS)

    The paper focuses on assessment of the capability of RELAP5 to predict natural circulation two-phase flow instability. The aim of this study is to identify needs for improvements of the code models. Experimental data from the low pressure CIRCUS facility was used for the code validation. The paper discusses the code validation procedure which combines separate and integral effect validation with the elements of the input errors propagation method for sensitivity analysis. The separate effect validation and 'transparent box' approach to the experimental system helps to identify main sources of experimental data uncertainty. The paper also discusses modifications provided in the new series of experiments to reduce the experimental uncertainty. Finally, the paper comes up with the conclusions about uncertainty in the RELAP5 prediction for different regimes of two-phase oscillatory flows in the CIRCUS facility and necessity for the models improvements. (author)

  20. Calculated thermal-hydraulic response for Semiscale Mod-3 Test S-07-6 using RELAP5: a new LWR system analysis code

    International Nuclear Information System (INIS)

    The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze the response of Semiscale Mod-3 Test S-07-6. The objective of Test S-07-6 was to provide reference data to evaluate LWR integral blowdown, refill, and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injected into the intact loop cold leg. The calculated test results using RELAP5 illustrate many of the nonequilibrium and nonhomogeneous aspects of the ECC injection which are not directly observable in the test data. These results also demonstrate the capability of the RELAP5 code and compare well with the test data fo break flow, pressure, temperature, and density throughout the Semiscale Mod-3 system. The periodic depletion and replenishment of ECC water in the downcomer shown in the test data is also shown in the calculation

  1. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    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. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs

  2. Peer review of RELAP5/MOD3 documentation

    International Nuclear Information System (INIS)

    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

  3. Thermal hydraulic analysis of ETRR-2 using RELAP5 code

    International Nuclear Information System (INIS)

    The present work was developed within the frame of the IAEA Coordinated Research Project 1496 ''Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors''. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.

  4. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    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. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs.

  5. Evaluation of the RELAP5/MOD3 multidimensional component model

    International Nuclear Information System (INIS)

    Accurate plenum predictions, which are directly related to the mixing models used, are an important plant modeling consideration because of the consequential impact on basic transient performance calculations for the integrated system. The effect of plenum is a time shift between inlet and outlet temperature changes to the particular volume. Perfect mixing, where the total volume interacts instantaneously with the total inlet flow, does not occur because of effects such as inlet/outlet nozzle jetting, flow stratification, nested vortices within the volume and the general three-dimensional velocity distribution of the flow field. The time lag which exists between the inlet and outlet flows impacts the predicted rate of temperature change experienced by various plant system components and this impacts local component analyses which are affected by the rate of temperature change. This study includes a comparison of two-dimensional plenum mixing predictions using CFD-FLOW3D, RELAP5/MOD3 and perfect mixing models. Three different geometries (flat, square and tall) are assessed for scalar transport times using a wide range of inlet velocity and isothermal conditions. In addition, the three geometries were evaluated for low flow conditions with the inlet flow experiencing a large step temperature decrease. A major conclusion from this study is that the RELAP5/MOD3 multidimensional component model appears to be adequately predicting plenum mixing for a wide range of thermal-hydraulic conditions representative of plant transients

  6. RLP5SPL: a conversion program from RELAP5 output data to SPL format data

    International Nuclear Information System (INIS)

    A conversion program RLP5SPL has been developed, which converts RELAP5 output data to SPL format data. It has functions to extract plot informations from RELAP5 restart file and convert them to SPL data format. After conversion of RELAP5 output data, it is easy to perform unit conversion, comparisons with other calculational data and/or experimental data within figures, and plotting various type of figures including a bird-eye view of three dimensional surface. (author)

  7. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 8. Analysis of the OECD MSLB Benchmark Exercise III Using Coupled Codes RELAP5/PARCS and TRAC-M/PARCS

    International Nuclear Information System (INIS)

    The OECD Nuclear Science Committee has released a set of computational benchmark problems for calculation of reactivity transients in pressurized water reactors (PWR). A main steam line break (MSLB) transient based on the Three Mile Island (TMI-1) PWR was developed to assess the capability of coupled neutronics and thermal-hydraulics codes to analyze complex transients having coupled core-plant interactions. The PWR MSLB accident scenario is characterized by a rupture in one of the main steam lines of the secondary system, leading to a sudden overcooling of the corresponding primary loop water. The overcooled moderator represents a positive reactivity insertion, which must be overcome by the control rods. Best-estimate modeling of this event requires three-dimensional (3-D) spatial kinetics because of space-time variations of the core power distribution arising from the asymmetric cooling of the core and from the scram of the reactor with the highest worth rod stuck out of the core. The benchmark was split into three separate exercises: a plant system model with point reactor kinetics, a spatial kinetics model of the core with the plant response modeled with time-dependent core thermal-hydraulic boundary conditions, and a plant system model with spatial kinetics model of the core. Results presented here are only for Exercise III of the benchmark with the cross-section set that leads to a return to power (RTP) during the transient. The work utilized the U.S. Nuclear Regulatory Commission (NRC) best-estimate thermal-hydraulic codes RELAP5 and TRAC-M coupled with the NRC neutronics code PARCS. The codes are coupled using a general interface (GI) incorporated into PARCS, which allows for the coupling to any thermal-hydraulics code. The thermal-hydraulics and neutronics codes are executed as separate processes with inter-process communication made possible through the use of message-passing protocols in the Parallel Virtual Machine (PVM) package. The spatial coupling of

  8. Peach Bottom Cycle 2 Low Flow Stability Tests analysis using RELAP5/PARCS

    International Nuclear Information System (INIS)

    Nowadays, the coupled codes technique, which consists in incorporating threedimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. So, in this work, the coupled codes technique using RELAP5/3.3-PARCS is applied to simulate stability transients in a BWR (Boiling Water Reactor). Validation has been performed against Peach Bottom-2 Low-Flow Stability Tests. In these transients dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein compared against the available experimental data. (author)

  9. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    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

  10. RELAP 5 Simulations of a hypothetical LOCA in Ringhals 2

    International Nuclear Information System (INIS)

    RELAP5 simulations of a hypothetical LOCA in Ringhals 2 were conducted in order to determine the sensitivity of the calculated peak cladding temperature (PCT) to Appendix K requirements. The PCT was most sensitive to the assumed model decay heat: Changing from the 1979 ANS Standard to 1.2 times the 1973 Standard increased the PCT by 70 to 100K. After decay heat, the two parameters which affected the PCT the most were steam generator heat transfer and heat transfer lockout. The PCT was not sensitive to the assumed pump rotor condition (locked vs coasting); nor was it sensitive to a modest amount (5 to 10%) of steam generator tube plugging. (author)

  11. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    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.

  12. ISP-46 analysis with RELAP5/SCDAPSIM computer code

    International Nuclear Information System (INIS)

    The thermal-hydraulic and severe accidents analysis code RELAP5/SCDAPSIM was used in the calculation of the Phebus FPT1 in-pile experiment. This experiment, carried out on 26 July 1996 in the Phebus facility, Cadarache, France, was chosen as the basis for the OECD International Standard Problem (ISP-46) exercise. Investigation of severe accidents phenomena like fuel degradation and hydrogen production was the objective of the ISP and of the presented analysis. The ISP was an open exercise, that is, all the relevant experimental results were available to the participants from the start. The FPT1 test bundle included 18 PWR fuel rods previously irradiated to a mean burnup of 23.4 GWd/tU, two instrumented fresh fuel rods and one silver-indium-cadmium control rod. The bundle was housed in an insulating shroud and introduced into the Phebus driver core which supplied the nuclear power. The fuel degradation phase of the test lasted about 5 hours during which the bundle was cooled by steam at pressure of about 2 bar with the mass flow rate varying between 0.5 g/s and 2.2 g/s, while the bundle nuclear power was being progressively increased from zero up to 36.5 kW. RELAP5/SCDAPSIM modelling of the Phebus facility and the main results, such as the temperature response of all rods and shroud, the oxidation and resulting hydrogen production, will be discussed and presented in this paper. The analysis of fuel rods degradation and problems related to SCDAPSIM underprediction of the amount of relocated fuel and cladding will also be covered. (author)

  13. RELAP5/MOD3 model and transport analyses for Maria research reactor in Poland

    International Nuclear Information System (INIS)

    RELAP5/MOD3 input data model of the Maria research reactor has been developed to provide the capability for the analysis of the reactor core under loss of flow and reactivity insertion transients. The model was qualified against the reactor data at steady state conditions and, additionally, against the existing reliable experimental data for the transient initiated by the reactor scram. The results obtained with the code agree well with the experimental data. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. The presented input data model should be treated as a first step for developing of the model including the whole primary cooling circuit of the reactor. (author)

  14. Mathematic preprocessor for RELAP5 code using Microsoft Excel; Pre-processador matematico para o codigo RELAP5 utilizando o Microsoft Excel

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Patricia Andrea

    2006-07-01

    Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a toll for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-Excel software. (author)

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

    International Nuclear Information System (INIS)

    This paper describes the on-going and finished parts of project: he external coupled RELAP5 DYN3D code. The RELAP5 is thermo-hydraulics code used for analysis of the thermohydraulics problems the nuclear facilities. The DYN3D is three-dimensional dynamic code used to calculate the dynamics processes the nuclear core. (author)

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

    International Nuclear Information System (INIS)

    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)

  17. Comparison of RELAP5/MOD2 and RELAP5/MOD3.1 calculating using FFT method

    International Nuclear Information System (INIS)

    In the paper described the comparison of the calculated results of two different versions of RELAP5 thermalhydraulic computer code. The scenario that all feedwater is lost during Small Break Loss of Coolant Accident (SBLOCA) was selected for comparison of the calculations. The break spectrum from 2.5 cm to 15.2 cm of equivalent diameter break size was analysed and compared using FFT method. The break was located in the cold leg. The methodology developed the University of Pisa (1) and at based on the Fast Fourier Transform (FFT) was applied to quantitatively assess and compare the calculations of small break loss of coolant accident. The analyses were performed at different steps during the transient. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Comparative assessment of coupled RELAP5/PARCS and DYN3D/RELAP5 codes against the Kozloduy-6 pump trip test

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y.; Grundmann, U.; Kliem, S.; Rohde, U. [Institute of Safety Research, FZR, Dresden (Germany); Bousbia Salahn, A. [Pisa Univ., DIMNP (Italy)

    2005-07-01

    The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate (BE) three-dimensional coupled code computational tools. Nowadays, this technique is extensively used since it allows decreasing conservatism in the calculation models by performing more realistic simulations based on a more precise consideration of multidimensional effects under complex transients in NPPs. This paper represents a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER-1000 pump trip test. The coupled RELAP5/3.3-PARCS/2.6 and DYN3D/3-RELAP5/3.3 code systems are used in simulations. The obtained results are assessed against experimental data and also through the code-to-code comparison. The DYN3D/RELAP5 computational model of VVER-1000 has been developed and adjusted for simulations with the parallel running scheme (PVM) of RELAP5/PARCS. Also, the macroscopic cross-section library used in the DYN3D/RELAP5 calculations has been adapted to meet the input requirements of PARCS. Prior to the test simulations, the RELAP5/PARCS model of the plant has been assessed in the stand-alone PARCS and RELAP5 test calculations. A reasonably good agreement between the experimental data and the calculated results is obtained. For the initial state, the observed discrepancies are mainly due to the absence of assembly discontinuity factor (ADF) correction and the evaluation of the Doppler feedback effect. During the transient, the deviations are mainly due to the combined effect of the measurement uncertainty in the control rod axial position and the estimation of the Doppler effect. (authors)

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

    International Nuclear Information System (INIS)

    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

  1. Validation of RELAP5/MOD3.2 model on trip off one main coolant pump for VVER 440/V230

    Energy Technology Data Exchange (ETDEWEB)

    Groudev, Pavlin [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, Antoaneta [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg

    2006-06-15

    This paper presents the validation of RELAP5/MOD3.2 model of the VVER 440 for Nuclear Power Plant (NPP) in the analysis of the following transient: 'Trip off one MCP'. This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with measurement transient data received from Kozloduy NPP Unit no. 4. The baseline input deck for VVER440 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 comparison between the RELAP5 calculations and the test data indicates a good agreement. This validation was possible through the participation of leading specialists from Kozloduy NPP and with the support of the Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

  2. High Flux Isotope Reactor system RELAP5 input model

    International Nuclear Information System (INIS)

    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

  3. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    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

  5. Peach bottom cycle 2 stability analysis using RELAP5/PARCS

    International Nuclear Information System (INIS)

    Boiling channels and systems may oscillate owing to the behaviour of the liquid-steam mixture used for removing the thermal power. A thermal-hydraulic system may be unstable under particular operating conditions. Two kinds of power oscillation have been observed in BWR cores. One is an in-phase (core-wide) and the other is an out-of-phase (regional) oscillation. Since the above feature can make detection more difficult, the latter oscillation is potentially more severe. The problem is well known since the design of the first BWR system. However, to improve the safety systems of these reactors, it is necessary to be able to detect in a reliable way these oscillations from the neutronic signals. The purpose of this work is to characterize the unstable behaviour of a BWR. Within this study, it has been performed a number of perturbation analysis. The coupled codes RELAP5-Mod3.3/PARCS have used for the simulation of the transients. Validation has been performed against Peach Bottom-2 Low-Flow Stability Test PT3. Three dimensional time domain BWR stability analysis were performed on test point 3 for the core wide oscillation mode. In this transient dynamically complex events take place, i.e., neutron kinetics is coupled with thermal-hydraulics and an in-phase oscillation has been developed. The calculated results are compared against the available experimental data. (author)

  6. Improvement of the RELAP5 subcooled boiling model for low pressure conditions

    International Nuclear Information System (INIS)

    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)

  7. Assessment of RELAP5-3D{copyright} using data from two-dimensional RPI flow tests

    Energy Technology Data Exchange (ETDEWEB)

    Davis, C.B.

    1998-07-01

    The capability of the RELAP5-3D{copyright} computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code`s logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved.

  8. PCRELAP5: data calculation program for RELAP 5 code; PCRELAP5: programa de calculo dos dados de entrada para o codigo RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre, Larissa Jacome Barros

    2016-07-01

    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. 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 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. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. 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 Calculo 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. In addition, 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. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Belchior Junior, Antonio; Andrade, Delvonei A.; Sabundjian, Gaiane; Macedo, Luiz A.; Angelo, Gabriel; Torres, Walmir M.; Umbehaun, Pedro E.; Conti, Thadeu N., E-mail: abelchior@ipen.br, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br, E-mail: lamacedo@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: tnconti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Bruel, Renata N. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

    2011-07-01

    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)

  10. 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.

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

    International Nuclear Information System (INIS)

    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)

  12. Simulation of condensation in a closed, slightly inclined horizontal pipe with a modified RELAP5 code

    OpenAIRE

    Szijártó, Rita; Freixa Terradas, Jordi; Prasser, Horst-Michael

    2014-01-01

    The performance of the RELAP5 thermal-hydraulic system code was analyzed in predicting very fast transient condensation processes in horizontal pipes. The code significantly underpredicted the heat transfer from the primary to the secondary side in case of rapid wall condensation process in the so called Inverse Edwards Pipe Experiment, where the condensation pipe was immerged in a cool water pool, and hot steam injection was performed into a pipe, which was closed on one side. The RELAP5 con...

  13. Modeling the thermal–hydraulic behavior of the reactor cavity cooling system using RELAP5-3D

    International Nuclear Information System (INIS)

    Highlights: • The RCCS complex geometry and heat transfer mechanisms were modeled with RELAP5-3D. • Code limitations were overcome by applying special heat structures modeling techniques. • The simulation results were found to be in good agreement with the experimental data. • RELAP5-3D was found to be an adequate tool for analysis of HTGR components. - Abstract: The Very High Temperature Gas-Cooled Reactor (VHTR), one of the six proposed designs for the next generation nuclear reactor, was conceived to achieve high temperatures to support industrial applications and power generation. Due to the high temperature reached during normal operation, the design included new passive safety systems. The Reactor Cavity Cooling System (RCCS) is a new passive safety system designed to remove the heat from the reactor cavity during normal operation (steady-state) and accident scenarios. Computational tools such as system codes have been selected to simulate the reactor system and, in particular, the new safety components. The capabilities of these codes are being investigated to verify their ability in predicting the phenomena involved in the RCCS operation during steady-state and accident conditions. A RELAP5-3D input model of a small scale water-cooled experimental facility was prepared to simulate steady-state. The simulation results were compared with data produced during the experimental steady-state run. The results obtained and presented in the paper showed a good agreement of the code prediction with the experimental data. The paper also provides a set of modeling techniques to overcome some of the limitations of the current version of the computer code in simulating complex geometries with combined heat transfer mechanisms in the reactor cavity of the VHTR

  14. Assessment of a RELAP5 model for the IPR-R1 Triga research reactor

    International Nuclear Information System (INIS)

    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. However, several current investigations have shown that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this way, as a contribution to the assessment of RELAP5/3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed by a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 kW. The reactor is located at the Nuclear Technology Development Centre (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 and also calculation data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code were considered in the process of the model validation. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual reactor behavior in good agreement with the available data. (author)

  15. Development and validation of coupled PARCS/RELAP5 model for Forsmark NPP at uprated power

    International Nuclear Information System (INIS)

    This paper gives an account of the development and validation of an up-to-date coupled neutronic/thermal-hydraulic model for the Swedish Forsmark boiling water reactor. The model will be used for analyses of the consequences of the planned power uprate from 2928 MWth to 3253 MWth. At first, the development of the PARCS and RELAP5 models are presented. On the neutronic side, cross-sections data was generated, allowing feeding PARCS with realistic data. This step was performed by converting the library data file from the power plant using the in-house cross-section interface code. The dependence of the material properties on history effects, burnup, and instantaneous conditions was accounted for, and the full heterogeneity of the core was thus taken into account. Each of the 676 fuel assemblies was modeled individually, while the 161 control rods were grouped into 6 different types. On the thermal-hydraulic side, the model consists of a model for the feedwater system, a model for the reactor vessel that include a model for the core channels, and a model for each of the four steam lines. The fuel assemblies were modeled as twelve flow channels in the core region. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The validation efforts were focusing on benchmarking the code capabilities against measured plant data, both under steady-state and transient conditions. The PARCS standalone model was validated against traversing in-core probe (TIP) measurements, taken at different burnup level with operating power varies from 108% (nominal level) to 120% (uprated level). The coupled PARCS/RELAP5 model was validated against an operational transient. For this validation task, the transient chosen was a turbine trip test, which was performed on May 6, 2013. Comparisons between calculated and measured parameters demonstrate that the coupled model was able to correctly represent the

  16. RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

    International Nuclear Information System (INIS)

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.

  17. RELAP5-3D results for phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW benchmark

    International Nuclear Information System (INIS)

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2. (authors)

  18. RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom

    2012-06-01

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.

  19. Analyses of PACTEL passive safety injection experiments with APROS, CATHARE and RELAP5 codes

    International Nuclear Information System (INIS)

    The European Commission fourth framework programme project 'Assessment of passive safety injection systems of advanced light water reactors' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of passive safety injection systems (PSIS) of advanced light water reactors (ALWRs) in small break loss-of-coolant accident (SBLOCA) conditions. The PSIS consisted of a core make-up tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the phenomena in the PSIS. (orig.)

  20. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    Energy Technology Data Exchange (ETDEWEB)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

  1. The assessment of RELAP5/MOD2 based on pressurizer transient experiments

    International Nuclear Information System (INIS)

    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)

  2. 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)

  3. Parametric study of different perturbations on Ringhals stability benchmark with RELAP5/PARCS

    International Nuclear Information System (INIS)

    The analysis of power instabilities were tackled many years ago developing new methodologies to model this phenomena. The mechanisms underlying the causes of the power oscillations in BWR are still under study, but its consequences are well known. The simulation of the instabilities using best-estimate codes is the aim of this work. Three dimensional time domain BWR stability analysis has been performed in Ringhals 1 NPP, using the coupled code RELAP5-MOD3.3/PARCS v2.7. In the simulation of instabilities, it is necessary to introduce some perturbations that make the power oscillate. In this work, the instabilities are induced by means of density perturbations using a new capability introduced in the neutronic code. The applied perturbation is based on the Lambda modes and their amplitudes. This new option permits the user to perturb the moderator density in each node at each time step. Using different amplitudes for the perturbation signal the user is able to perform a complete stability analysis studying the resulting power oscillations. (author)

  4. 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.

  5. ROSA-IV LSTF 5% cold leg break analysis using the RELAP5/Mod2 code

    International Nuclear Information System (INIS)

    This paper presents the results of the OECD International Standard Problem calculation no. 26 (ISP-26), performed with the RELAP5/Mod2 code (frozen version 36.06 for the CRAY-X/MP) at the Paul Scherrer Institute (PSI) and analyses their comparison with the 5% cold leg break test (run SB-CL-18) conducted on the Large Scale Test Facility (LSTF) of the ROSA-IV program. The principal objective of the present calculation is the analysis of the simulation capability of the code in regards to the main phenomena occurring during a typical Small Break Loss of Coolant Accident (SB-LOCA), such as: natural circulation heat removal; liquid holdup in the up- and downflow sides of the steam generator U-tubes in combination with a reflux condenser mode heat removal, core upper plenum pressure build-up causing manometric liquid level unbalance between core and downcomer; loop seal clearing concurrent with core liquid depression leading to core heat-up; core level recovery after loop seal clearing; vessel inventory boil-off leading to second core uncovering; core reflooding due to accumulator injection intervention. 7 refs., 15 figs., 2 tabs

  6. An assessment of RELAP5/MOD2 applicability to loss-of-feedwater transient analysis in a Babcock and Wilcox reactor plant

    International Nuclear Information System (INIS)

    The applicability and scaling capability of RELAP5/MOD2 when applied to a Babcock and Wilcox (B and W) loss-of-feedwater transient is assessed using a code applicability methodology. A loss-of-feedwater test with a feed-and-bleed recovery was selected from the once-through integral system (OTIS) test data as a reference transient. Nondimensional comparisons are made between code assessment calculations and code applications calculations using computer code models scaled according to scaling criteria derived from the work of Ishii and others. The results indicate that RELAP5/MOD2 can scale the phenomena observed in the experiment and that the code is applicable for transients for which phenomena are within this envelope. The results also demonstrate the usefulness of the code applicability methodology for interpreting and verifying code calculations. 21 refs., 59 figs., 12 tabs

  7. Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS v2.7 coupled code

    International Nuclear Information System (INIS)

    To efficiently characterize realistic transients, as the Reactivity Insertion Accidents (RIA), using coupled neutronic-thermal-hydraulic 3D best estimate system codes, like RELAP5/PARCS v2.7 coupled code, it is necessary to introduce some improvements in simulations by adding the capability of control rod movement and boron injection by means of RELAP5 control variables, with the aim of being able to analyze dynamically asymmetric transient accidents in a nuclear power reactor, like RIA, reproducing all control systems present in commercial reactors. In actual neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. With these modifications, control rods can be moved automatically, activated by the RELAP code control system, and also they can depend on signals related to the reactor activity, like pressure, fuel temperature or moderator temperature, etc., improving the realism of the calculation and widening the simulation possibilities. RELAP5 calculates the boron concentration in each node of the channels representing the reactor core, sending this information to the PARCS neutronic code. The environment of work chosen have been the graphical environment of programming Compaq Visual Fortran 6.6A (CVF 6.6A). The fundamental reasons have been, on the one hand the facility of

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

    International Nuclear Information System (INIS)

    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.

  9. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

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

    International Nuclear Information System (INIS)

    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

  12. Implementation of DOWTHERM A Properties into RELAP5-3D/ATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Richard L. Moore

    2010-04-01

    DOWTHERM A oil is being considered for use as a heat transfer fluid in experiments to help in the design of heat transfer components for the Next Generation Nuclear Plant (NGNP). In conjection with the experiments RELAP5-3D/ATHENA will be used to help design and analyzed the data generated by the experiments. Inorder to use RELAP5-3D the thermophysical properties of DOWTHERM A were implemented into the fluids package of the RELAP5-3D/ATHENA computer propgram. DOWTHERM A properties were implemented in RELAP5-3D/ATHENA using thermophysical property data obtain from a Dow Chemical Company brochure. The data were curve fit and the polynomial equations developed for each required property were input into a fluid property generator. The generated data was then compared to the orginal DOWTHERM A data to verify that the fluid property data generated by the RELAP5-3D/ATHENA code was representitive of the original input data to the generator.

  13. RELAP5/MOD3.2 investigation of a WWER-440 steam generator header cover lifting

    International Nuclear Information System (INIS)

    This report discussed the results of the thermal-hydraulic analysis of an accident at the Rovno NPP. The accident was caused by primar to secondary reactor coolant leakage as a result of full and partial steam generators header covers lifting. The initiating event is full hot collector cover lifting in one of the sixth steam generators (SG) with equivalent diameter 107 mm. Hot collector cover lifting in other three SGs - 1, 3 and 4 follows this event. Such accident provide a direct release path for contaminated primary coolant to the environment via the secondary side. RELAP5/MOD3.2 computer code has been used to simulate the transient at the Rovno WWER-440 NPP. A model of the Rovno unit 1 has been developed based on the RELAP5/MOD3.2 thermal-hydraulic code at the National Taras Shevchenko University of Kiev and has been given to the Institute for Nuclear Research and Nuclear Energy (INRNE) staff for performing of thermal-hydraulic analyses. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of WWER-440 mentioned above against the experimental transient data received from the Rovno WWER-440 NPP, unit 1. The results of this investigation provide an integrated evaluation of the complete RELAP5 WWER-440/V213 model. As it seen from the results RELAP5 predict correctly the behaviour of main plant parameters. (authors)

  14. Transient validation of RELAP5 model with the DISS facility in once through operation mode

    Science.gov (United States)

    Serrano-Aguilera, J. J.; Valenzuela, L.

    2016-05-01

    Thermal-hydraulic code RELAP5 has been used to model a Solar Direct Steam Generation (DSG) system. Experimental data from the DISS facility located at Plataforma Solar de Almería is compared to the numerical results of the RELAP5 model in order to validate it. Both the model and the experimental set-up are in once through operation mode where no injection or active control is regarded. Time dependent boundary conditions are taken into account. This work is a preliminary study of further research that will be carried out in order to achieve a thorough validation of RELAP5 models in the context of DSG in line-focus solar collectors.

  15. Comparison between RELAP5/MOD1 and TRAC-PD2 codes in the CANON experience

    International Nuclear Information System (INIS)

    The present work reports comparisons between experimental and theoretical data done with the RELAP5/MOD1 and TRAC-PD2 codes, with particular emphasis on RELAP5/MOD1 code run with basic experimental data from the CANON depressurization simulation. This experiment simulates a Loss of Primary Coolant Accident due to a Large Rupture - LOCA, through the depressurization of a horizontal tube containing water with the instantaneous break of one side of the tube where measurements of pressure, temperature and void fraction are taken during the transient. The results of this comparison show that RELAP5/MOD1 code predict more satisfactorily the time dependent behavior of the pressure and void fraction than TRAC-PD2 code for several initial condition considered in the CANON experiment. (author)

  16. RELAP5 assessment: LOFT small-break L3-6/L8-1

    International Nuclear Information System (INIS)

    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 LWR's during accident and off-normal conditions. The RELAP5 code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, a small break transient and subsequent partial core uncovery transient performed at the LOFT facility have been analyzed. The results show that RELAP5/MOD1 does very well on predicting the qualitative behavior for this small break experiment, although there are a number of quantitative disagreements

  17. RELAP5 simulation for one and two-phase natural circulation phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de; Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Castro, Alfredo Jose Alvim de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: gdjian@ipen.br; delvonei@ig.com.br; umbehaun@ipen.br; wmtorres@ipen.br; Braz Filho, Francisco A.; Borges, Eduardo Madeira [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados]. E-mails: eduardo@ieav.cta.br; fbraz@ieav.cta.br; Belchior Junior, Antonio; Rocha, Ricardo Takeshi Vieira da [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mails: belchior@bol.com.br; rtvrocha@uol.com.br; Damy, Osvaldo Luiz Almeida; Torres, Eduardo [Universidade de Sao Paulo (USP), SP (Brazil). Escola Politecnica]. E-mails: osvaldo.damy@poli.usp.br; etorres@pac.ind.br

    2007-07-01

    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)

  18. RELAP5 simulation for one and two-phase natural circulation phenomenon

    International Nuclear Information System (INIS)

    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)

  19. Analysis of high pressure sequence for an integral reactor using SCDAP/RELAP5

    International Nuclear Information System (INIS)

    The high RCS (Reactor Coolant System) pressure sequence of a TLOFW (Total Loss Of Feed Water) for an integral reactor was analyzed in order to obtain an overall insight into a severe accident progression from an initiation of accident to the reactor vessel failure in detail by using the SCDAP/RELAP5 computer code. The present results were compared with the typical PWR (Pressurized Water Reactor) results. The SCDAP/RELAP5 results have shown that the internal structure in the integral reactor influenced the in-vessel melt progression and the reactor vessel failure time. For this reason, the SCDAP/RELAP5 results with no simulation of the internal structures were very similar to the typical PWR results. (authors)

  20. Development of Ignalina NPP RBMK-1500 reactor RELAP5-3D model

    International Nuclear Information System (INIS)

    This paper deals with the development of an integrated thermal-hydraulics-neutronics model for RBMK-1500 reactors for the analysis of specific plant transients in which the neutronic response of the core is important. A successful best estimate coupled RELAP5-3D model of Ignalina nuclear power plant (NPP) has been developed. The validation of the thermal-hydraulic model has been performed using operational transients from Ignalina NPP. The results of the calculations obtained with the RELAP5-3D model compare reasonably with the real plant data. The RELAP5-3D nodal kinetics model provides reasonable agreement with Ignalina NPP reactor power and coolant density profiles. The eigenvalue is close to unity, indicating that reasonable values are calculated for the neutron fluxes

  1. 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.

  2. RELAP5 assessment: LOFT turbine trip L6-7/L9-2

    International Nuclear Information System (INIS)

    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 is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, a turbine trip rapid cooldown transient performed at the LOFT test facility has been analyzed. The results show that RELAP5/MOD1 can predict the experimental behavior of LOFT test L6-7/L9-2 in detail. However, careful selection of modeling options and adjustment of boundary conditions within the experimental uncertainties is required

  3. Assessment and modifications of the post-CHF wall heat transfer packages of RELAP5/MOD2.5 and RELAP5/MOD3

    International Nuclear Information System (INIS)

    During the last few months, considerable effort has been spent on assessing the post-CHF wall heat transfer package of RELAP5/MOD3/v7j (R5M3) as well as on investigating the effect of some model differences between this code and its predecessor, RELAP5/MOD2.5 (R5M2). In this work, the authors outline the problems associated with the post-CHF wall heat transfer models and logic of R5M3 (which are partly responsible for the totally unphysical code predictions during reflooding), the main differences between R5M2 and R5M3 and the author shows that by implementing in both codes a physically realistic and sound wall-to-liquid heat transfer model, one can predict very well experimental results obtained in separate-effect bottom flooding tests

  4. Assessment of RELAP5/MOD3.1 for gravity-driven injection experiment in the core makeup tank of the CARR Passive Reactor (CP-1300)

    International Nuclear Information System (INIS)

    The objective of the present work is to improve the analysis capability of RELAP5/MOD3.1 on the direct contact condensation in the core makeup tank (CMT) of passive high-pressure injection system (PHPIS) in the CARR Passive Reactor (CP-1300). The gravity-driven injection experiment is conducted by using a small scale test facility to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes. It turns out that the larger the water subcooling is, the more initiation of injection is delayed, and the sparger and the natural circulation of the hot water from the steam generator accelerate the gravity-driven injection. The condensation modes are divided into three modes: sonic jet, subsonic jet, and steam cavity. RELAP5/MOD3.1 is chosen to evaluate the cod predictability on the direct contact condensation in the CMT. It is found that the predictions of MOD3.1 are in better agreement with the experimental data than those of MOD3.0. From the nodalization study of the test section, the 1-node model shows better agreement with the experimental data than the multi-node models. RELAP5/MOD3.1 identifies the flow regime of the test section as vertical stratification. However, the flow regime observed in the experiment is the subsonic jet with the bubble having the vertical cone shape. To accurately predict the direct contact condensation in the CMT with RELAP5/MOD3.1, it is essential that a new set of the interfacial heat transfer coefficients and a new flow regime map for direct contact condensation in the CMT be developed

  5. Input Calibration and Validation of RELAP5 Against CIRCUS-IV Single Channel Tests on Natural Circulation Two-Phase Flow Instability

    Directory of Open Access Journals (Sweden)

    Viet-Anh Phung

    2015-01-01

    Full Text Available RELAP5 is a system thermal-hydraulic code that is used to perform safety analysis on nuclear reactors. Since the code is based on steady state, two-phase flow regime maps, there is a concern that RELAP5 may provide significant errors for rapid transient conditions. In this work, the capability of RELAP5 code to predict the oscillatory behavior of a natural circulation driven, two-phase flow at low pressure is investigated. The simulations are compared with a series of experiments that were performed in the CIRCUS-IV facility at the Delft University of Technology. For this purpose, we developed a procedure for calibration of the input and code validation. The procedure employs (i multiple parameters measured in different regimes, (ii independent consideration of the subsections of the loop, and (iii assessment of importance of the uncertain input parameters. We found that predicted system parameters are less sensitive to variations of the uncertain input and boundary conditions in high frequency oscillations regime. It is shown that calculation results overlap experimental values, except for the high frequency oscillations regime where the maximum inlet flow rate was overestimated. This finding agrees with the idea that steady state, two-phase flow regime maps might be one of the possible reasons for the discrepancy in case of rapid transients in two-phase systems.

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    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

  9. An assessment of RELAP5-3D using the Edwards-O'Brien Blowdown problem

    International Nuclear Information System (INIS)

    The RELAP5-3D (version bt) computer code was used to assess the United States Nuclear Regulatory Commission's Standard Problem 1 (Edwards-O'Brien Blowdown Test). The RELAP5-3D standard installation problem based on the Edwards-O'Brien Blowdown Test was modified to model the appropriate initial conditions and to represent the proper location of the instruments present in the experiment. The results obtained using the modified model are significantly different from the original calculation indicating the need to model accurately the experimental conditions if an accurate assessment of the calculational model is to be obtained

  10. Vector and parallel processing of the nuclear reactor transient analysis code RELAP5

    International Nuclear Information System (INIS)

    An experiment of vector processing and multi-tasking of nuclear reactor transient analysis code RELAP5 has been made at Japan Atomic Energy Research Institute. Vector processing and multi-tasking of the RELAP5 were achieved by using the independency of the spatial meshes. The vectorization ratio is 83%. The performance ratio in the vector mode to that in the scalar mode is about 3 on the FACOM VP-100. For multi-tasking, the spatial meshes are halved and each group of meshes is processed on different processor. The effect of multi-tasking was estimated from the CPU time on the FACOM VP-100, which is single processor system

  11. Investigation of Solution Methods Suitable for Modelling Steam Collapse and Pressure Peaks Using Relap5

    International Nuclear Information System (INIS)

    In this report an investigation of solution methods suitable for modeling steam collapse and pressure peaks using RELAP5 (modification 3.3, patch 03) is presented. Simulations using RELAP5 can deviate much from the reality as steam collapse occurs in the system. Steam collapse results in high and sudden pressure peaks and from a safety point of view it is essential that the pressure amplitudes are overestimated in the simulations. The uncertainty in the modeling results using RELAP5 can cause problems, for example when dynamic loads for nuclear power plants are calculated. There are today no clear instructions on which settings should be used in RELAP5 during such conditions or thorough investigation to validate the deviation of the results from the actual values. This has imposed problems in validating that load calculations using RELAP5 when steam collapse occurs are correct. The purpose of this project is therefore to develop a solution method suitable for RELAP5 simulations when steam collapse occurs to better simulate the actual force on the system. In the future, this developed solution method can be used when no experimental data is available and this work can be used as a reference as how to handle certain types of steam collapse and as a validation that this solution method does not result in an underestimation of the pressure peaks. Experimental data is available for three different setups, where one of the experiments was performed at the auxiliary feed water system at Ringhals 1. The method used for obtaining a suitable solution method, is to compare the experimental data to simulations performed using different solution approaches in RELAP5. Three main changes to the solution approach are investigated: activation of the equilibrium solution and homogenous solution and deactivation of the critical flow model. In addition to these changes some further solution approaches, not anticipated to affect the results in a higher degree, are investigated. The

  12. Assessment and improvement of condensation model in RELAP5/MOD3

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Thermal hydraulic analysis of the multipurpose research reactor RMB using a RELAP5 model

    International Nuclear Information System (INIS)

    The Multipurpose Brazilian Reactor (RMB) will be an open pool multipurpose research reactor using low enriched uranium fuel (LEU). This paper presents the RMB nodalization and the first thermal hydraulic results of steady state calculations using the RELAP5-MOD3.3 code. Several current investigations have shown that RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions in spite of such code was initially projected to studies of commercial nuclear power plants. (author)

  16. High-fidelity, real-time simulation with RELAP5/NESTLE

    International Nuclear Information System (INIS)

    Recent advances in the computational speed of engineering workstations have enabled the development of a real-time version of the RELAP5 nuclear plant simulation code with laboratory discretionary research and development funding. In addition, the Idaho National Engineering Laboratory (INEL) is also funding the development of an enhanced real-time version of the existing three-dimensional nodal neutron kinetics package via its University Research Consortium (URC) at Purdue University and North Carolina State University (NCSU). This paper focuses on the enhancements to RELAP5 to achieve real-time performance

  17. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  18. SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

    International Nuclear Information System (INIS)

    Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. Currently, the COUPLE model has no capability to model the movement of material that melts within a matrix of porous material. The COUPLE model also does not have the capability to model the movement of liquefied core plate material that slumps onto a porous debris bed in the lower head. In order to advance beyond the assumption the liquefied material always remains stationary, designs are developed for calculations of the movement of liquefied material through the interstices in a matrix of porous material. Correlations are identified for calculating the permeability of the porous debris and for calculating the rate of flow of liquefied material through the interstices in the debris bed. Correlations are also identified for calculating the relocation of solid debris that has a large amount of cavities due to the flowing away of melted material. Equations are defined for calculating the effect on the temperature distribution in the debris bed of heat transported by moving material and for changes in effective thermal conductivity and heat capacity due to the movement of material. The implementation of these models is expected to improve the calculation of the material distribution and temperature distribution of debris in the lower head for cases in which the debris is porous and liquefied material is present within the porous debris

  19. ISP33 Natural single and two phase flow in PACTEL blind calculation using RELAP5/Mod3

    International Nuclear Information System (INIS)

    The OECD/CSNI standard problem ISP33 aimed at adding information about natural circulation phenomena which are of particular importance during shut down of a reactor. The experiment performed in the Finnish PACTEL facility was carried out at a constant low core power simulating about 3.4% residual heat. The coolant inventory was stepwise reduced at 900 s intervals by about 10%. Thus effects of the various contents of steam in the primary could be studied during time intervals with semi-steady natural one- and two-phase circulation driven by the core rest heating. Results obtained in a blind calculation compare generally well with the experiment which shows that the RELAP5/Mod3 code is capable of predicting natural circulation phenomena. Exceptions are pressure peaks after the second drain and delayed final core heat-up which could be explained by inadequacies in certain measured data such as the core power. 10 refs

  20. RELAP5 code validation using a medium-size break LOCA experiment at the PMK-2 test facility

    International Nuclear Information System (INIS)

    For the analyses of loss of coolant accidents (LOCA) the thermohydraulic computer code capabilities for eastern-type reactors like VVER-440 must be validated by pre- and post test calculations of suitable experiments. Such experiments are performed on PMK-2 integral-type test facility in KFKI Atomic Energy Research Institute, Budapest, which is a volume-scaled model of the primary and secondary system of the Paks Nuclear Power Plant. One of these experiments is the pressuriser surge line break which correspond to a 22% leak. The most important phenomena of the experiment are the behavior of hot leg loop seal and the core dry-out with refill-reflood. Posttest calculations were performed by use of the code version RELAP5/mod.3.2. The results of the calculation and experiment are compared. The code properly simulate the analyzed transient.(author)

  1. Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code

    Science.gov (United States)

    De Rosa, F.; Lombardo, C.; Mascari, F.; Polidori, M.; Chiovaro, P.; D'Amico, S.; Moscato, I.; Vella, G.

    2014-11-01

    The present paper deals with the investigation of the evolution and consequences of a Station Black-Out (SBO) initiating event transient in the SPES3 facility [1]. This facility is an integral simulator of a small modular reactor being built at the SIET laboratories, in the framework of the R&D program on nuclear fission funded by the Italian Ministry of Economic Development and led by ENEA. The SBO transient will be simulated by using the RELAP5 and TRACE nodalizations of the SPES3 facility. Moreover, the analysis will contribute to study the differences on the code predictions considering the different modelling approach with one and/or three-dimensional components and to compare the capability of these codes to describe the SPES3 facility behaviour.

  2. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2002-01-01

    Full Text Available This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to simulate the interactions between the generated steam line pressure wave propagation and the instantaneous core void distribution. An overall comparison shows good agreement between the code calculations and the experimental data. A series of sensitivity analyses were also performed in order to assess the code prediction features, as well as to identify uncertainties related to the adopted thermalhydraulic parameters used for the plant modelisation.

  3. Sub-channel analysis by RELAP5 system code of boil-off experiment (Test 5002) with NEPTUN facility

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A. [Pennsylvania State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, Pennsylvania (United States)]. E-mail: axp46@psu.edu; Bousbia Salah, A.; D' Auria, F. [Univ. of Pisa, Dipartimento di Ingegneria Meccanica, Nucleare d della Produzione, Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; f.dauria@ing.unipi.it

    2004-07-01

    This paper presents the results of RELAP5/Mod3.2 system thermalhydraulic code using the sub-channel analysis approach in predicting the NEPTUN separate effect boil off experiments. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the NEPTUN low pressure test N{sup o}5002 has been considered. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory and demonstrate, as well, the reasonable success of the 'sub-channel analysis' approach adopted in the present context for a system thermalhydraulic code. (author)

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

    International Nuclear Information System (INIS)

    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.

  5. 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.

  6. Employing of RELAP-5 code for LBB (leak-before-break) deterministic analysis in the Ignalina NPP

    International Nuclear Information System (INIS)

    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 a computer program that predicts the leakage rate for cracked pipes in NPP. As this program calculates only water leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. The RELAP5 (reactor excursion and leak analysis program) can model transients in light water reactors (LWR) systems, such as loss of coolant, operational transients, anticipated transients without scram (ATWS), e.g. loss of feed-water, loss of offsite power, station blackout, and turbine trip. The RELAP5 code employs hydrodynamic, heat structure and reactor kinetics models with control and trip systems and time step control. If crack opening area (COA) is sufficiently big, then crack can be modeled as narrow pipe. The pipe cross-section shape can be assumed as round, rectangular and elliptical. This paper shows, that RELAP5 code could be employed for calculations of coolant discharge through cracks. For model verification a comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows that calculated RELAP5 and SQUIRT results compare favourably with experimental data. It means, that RELAP5 model is suitable for calculations of leak through through-wall cracks in pipes. (A.C.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonella L., E-mail: lombardicosta@gmail.co [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Reis, Patricia Amelia L., E-mail: patricialire@yahoo.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Pereira, Claubia, E-mail: claubia@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Veloso, Maria Auxiliadora F., E-mail: dora@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN, Av. Antonio Carlos, 6627, Campus UFMG, Belo Horizonte (Brazil); Soares, Humberto V., E-mail: betovitor@ig.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2010-06-15

    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.

  8. Development of LabVIEW web-based simulator for RELAP5

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  10. 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)

  11. Recent Hydrodynamics Improvements to the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

    The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model.

  12. Recent Hydrodynamics Improvements to the RELAP5-3D Code

    International Nuclear Information System (INIS)

    The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model

  13. RELAP5/MOD3.3 analysis of reactor trip event in nuclear power plant

    International Nuclear Information System (INIS)

    Measured plant data from various abnormal events or incidents are of great importance for assessing large system thermal-hydraulic computer codes like RELAP5. In the present study the reactor trip, which occurred at Krsko Nuclear Power Plant (NPP) on April 10, 2005, has been analyzed. The purpose of the analysis was to assess the RELAP5/MOD3.3 Patch 03 computer code against plant measured data and validate the RELAP5 input model for Krsko NPP, which is a two-loop Westinghouse pressurized water reactor. The RELAP5 input model delivered by Krsko NPP was used. The event analyzed was a malfunction, which occurred during a power reduction sequence when regular periodic testing of the turbine valves was performed. This caused plant trip. The calculation agrees very well with the plant measured data when operator actions are modelled properly. It was found out that the long term transient evolution is very sensitive to the steam flows from the steam generators after the reactor trip and only proper modelling of these flows gives good quantitative agreement. (author)

  14. RELAP5 two-phase fluid model and numerical scheme for economic LWR system simulation

    International Nuclear Information System (INIS)

    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

  15. Application on the RELAP5 computer code to the WWER reactor type safety analysis

    International Nuclear Information System (INIS)

    The development and brief description are given of the RELAP5/MOD2 code and the history of its use at the Nuclear Research Institute (NRI), Rez, is presented. Information is given on some test case analysis performed in the NRI to study the LOCA phenomena on the WWER-440/213 reactor. (Z.S.) 7 figs., 6 refs

  16. 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)

  17. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    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 code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  20. Power loop modeling and simulation using LabVIEW coupled with RELAP5

    Science.gov (United States)

    Pack, Joshua C.

    The purpose of this thesis is to provide an additional tool to researchers and system analysts for use in simulation, testing, and development of the secondary loop of a PWR nuclear power plant. This new tool is a coupling of LabVIEW and RELAP5 that has been created by using each code to model half of a PWR. By taking advantage of the strengths of both programs, a more powerful, adaptable, and user friendly system model is developed that links directly to the instrumentation of the system. This work includes the development of the LabVIEW secondary loop model, the coupling methods for linking the two software packages, and a comparison of the secondary loop outputs to typical RELAP5 outputs as well as a third party source.

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

    International Nuclear Information System (INIS)

    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

  2. Assessment of PWR Steam Generator modelling in RELAP5/MOD2

    International Nuclear Information System (INIS)

    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

  3. Heat Transfer Boundary Conditions in the RELAP5-3D Code

    International Nuclear Information System (INIS)

    The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones

  4. Using code RELAP5 and RELAP/SCADAPSIM for research reactor LVR-15 and experimental equipments

    International Nuclear Information System (INIS)

    This paper describes spheres of using code RELAP5/3.2.2 and RELAP/SCADAPSIM by Reactor Services Division of Nuclear Research Institute in the Czech Republic. RELAP5/3.2.2 and RELAP/SCADAPSIM are used for the assessment of transient analysis of research reactor LVR-15 and for simulation of thermal-hydraulic conditions of high-pressure and high-temperature in-pile water loop facilities for testing of materials under various conditions of LWRs. Recently analysis are also made for the in-pile helium loop facility build with the aim to simulate thermal-hydraulic conditions and structure materials behaviour under VHTR conditions. (author)

  5. Evaluation of validity of the RELAP5/MOD3 flow regime map for horizontal tubes

    International Nuclear Information System (INIS)

    RELAP5/MOD3 code was developed for western type power water reactors with vertical steam generators. Thus, this code should be validated also for VVER design with horizontal steam generators. The validation work, which has been started in Lappeenranta University of Technology (LUT), has already shown some weaknesses of the code. For example the flow inside a steam generator horizontal tube in some accident cases is not correctly modelled by the code. It may be the result of erroneous prediction of the flow regime. The aim of the 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 tubes. (18 refs.)

  6. Relap5 based subchannel analysis for licensing of Atucha2 NPP

    International Nuclear Information System (INIS)

    The departure from nucleate boiling ratio (DNBR) is one of the figures of merit which are typically presented in the chapter accident analyses of a final safety analysis report. In Argentina, like in other countries, the owner of a NPP has to demonstrate that for transients, which are expected to happen during the lifetime of a NPP, dryout or film boiling can be excluded with 95% probability for a 95% tolerance interval. A precise method to evaluate DNBR leads to a smaller tolerance interval, and helps to avoid unnecessary conservatism in the evaluation. This paper presents DNBR evaluation for the Atucha 2 NPP (CNA2) using Relap5. CNA2 is a heavy water cooled, heavy water moderated pressurized water reactor with vertical coolant channels. It is shown that the results and uncertainties of a Relap5 subchannel analysis are comparable with those that are typically achieved by specialized subchannel codes. (author)

  7. Heat Transfer Boundary Conditions in the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-01

    The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.

  8. A new assessment of RELAP5-3D using a General Electric level swell problem

    Energy Technology Data Exchange (ETDEWEB)

    Aumiller, D.L.; Tomlinson, E.T.; Clarke, W.G.

    2000-09-01

    The RELAP5-3D (version bt) computer program was used to assess a GE level swell experiment. The primary goal of the new assessment models was to faithfully represent the experimental facility and instrumentation. In developing the new models, a non-physical representation of the vessel heads in a previous assessment was found. This distortion resulted in predictions that closely matched the experimental data, but were in error. The new assessment also highlighted an instability in the calculation of interfacial drag. To explore this issue, analyses were performed using three different interfacial drag correlations appropriate for large diameter pipes and/or vessels. The results of this study show that the Kataoka-Ishii correlation, which is currently used in RELAP5-3D, compares most favorably with the experimental data. Additionally, a numerical instability was uncovered with the analysis performed using the Gardner correlation and was traced to the calculation of bubble diameter in the bubbly flow regime.

  9. Simulation of SBLOCA based on an Improved Choked Flow Model for RELAP5/MOD3 Code

    International Nuclear Information System (INIS)

    This paper is a continuation of the present author's previous publication dealing with a new choked flow model for two-phase flow. The model based on a hyperbolic one-dimensional two-fluid model, where in the momentum equations the terms representing the interfacial pressure difference has been included in lieu of the virtual mass force terms. The new choked flow model is an improvement upon the choked flow model of the current RELAP5/MOD3 code, which itself is based on the Trapp-Ransom method. The author compares the predictions of this improved model with Trapp-Ransom model and Henry-Fauske model, for an assumed flow in a vertical pipe. The author simulates a typical PWR system with a hypothetical SBLOCA as well, and compares the system behaviors predicted by RELAP5/MOD3, based on the aforementioned choked flow models. He shows that the improved choked flow model leads to better predictions

  10. A simulation of steam generator tube rupture accident by safety analysis code RELAP5/MODI

    International Nuclear Information System (INIS)

    Steam-generator-tube-rupture accident occurred at Prairie Island unit 1 is simulated using the RELAP5/MOD1 code which has been developed as a best-estimate safety analysis code for light water reactors. The purpose of the simulation is to examine its capacity as a tool of obtaining high-quality and verified data base needed for developing diagnostic techniques of nuclear power plants. The simulation is conducted until 3200 seconds after the tube rupture. The simulation results agrees fairly well with both the plant records and the RETRAN-02 simulation results conducted at Japan Atomic Energy Research Institute, and it is concluded that the RELAP5/MOD1 code is effective to simulate the overall plant behavior during the accident, although several items remain for future improvement. (author)

  11. Analysis of postulated loss of coolant accidents on Brazilian Multipurpose Reactor using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Humberto Vitor; Costa, Antonella Lombardi; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Reis, Patricia Amelia de Lima, E-mail: hvs@cdtn.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Aronne, Ivan Dionysio, E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-07-01

    The Brazilian Multipurpose Reactor (RMB) is currently being projected and several analyses are being carried out. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly with planar plates. RMB will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of RMB using the RELAP5 model and also three cases of loss of coolant accident (LOCA), in the reactor and service polls cooling system (RSPCS) inlet and two cases in the primary coolant system (PCS), inlet and outlet. In both cases the coolant pool level decreased until 7 m, keeping the core covered by water, but in different times. Natural circulation mode was established in the reactor pool and consequently the decay heat was removed keeping the integrity of the fuel elements. Keywords: Research reactor, LOCA, RELAP5. (author)

  12. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    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.

  13. Plant application uncertainty evaluation of LBLOCA analysis using RELAP5/MOD3/KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chung, Bub Dong; Hwang, Tae Suk; Lee, Guy Hyung; Chang, Byung Hoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the revised ECCS rule has been developed and this report describes the application of new REM to large break LOCA. A computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/ MOD3.1 was used as the best estimated code for the analysis and Kori unit 3 and 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by using Monte-Carlo sampler for each response surface. This study shows that plant application uncertainty can be quantified and demonstrates the applicability of the new realistic evaluation methodology. (Author) 29 refs., 40 figs., 8 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  15. Analysis of postulated loss of coolant accidents on Brazilian Multipurpose Reactor using RELAP5

    International Nuclear Information System (INIS)

    The Brazilian Multipurpose Reactor (RMB) is currently being projected and several analyses are being carried out. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly with planar plates. RMB will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of RMB using the RELAP5 model and also three cases of loss of coolant accident (LOCA), in the reactor and service polls cooling system (RSPCS) inlet and two cases in the primary coolant system (PCS), inlet and outlet. In both cases the coolant pool level decreased until 7 m, keeping the core covered by water, but in different times. Natural circulation mode was established in the reactor pool and consequently the decay heat was removed keeping the integrity of the fuel elements. Keywords: Research reactor, LOCA, RELAP5. (author)

  16. 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)

    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)

  17. TMI-1 MSLB coupled 3-D neutronics/thermal hydraulics analysis: application of RELAP5-3D and comparison with different codes

    International Nuclear Information System (INIS)

    A comprehensive analysis of the double ended Main Steam Line Break (MSLB) accident assumed to occur in the Babcock and Wilcox nuclear power plant of Three Miles Island Unit 1 (TMI-1) has been carried out of the University of Pisa in co-operation with the University of Zagreb and the Texas A and M University. The overall activity has been completed within the framework of the participation in the OECD-CSNI/NSC (Committee on the Safety of Nuclear Installations - Nuclear Science Committee) 'PWR MSLB Benchmark'. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: RELAP5/MOD3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code; RELAP5/MOD3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code; RELAP5/3D, coupled with the 3-D neutron kinetics Nestle code. Boundary and initial conditions of the system including those relevant to the fuel status, have been supplied by Pensilvania State University that had a co-operation GPU (the utility, owner of TMI) and NRC (US Nuclear Regulatory Commission). The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 're-criticality' or 'return-to-power' whose magnitude is largely affected by boundary and initial conditions. The comparison among the results obtained by adopting the same thermalhydraulic nodalization and the different 'coupled' code version is discussed in the present document. (author)

  18. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  19. RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

    OpenAIRE

    Borut Mavko; Andrej Prošek

    2010-01-01

    To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA). The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA). For calculations the RELAP5/MOD3.3 best ...

  20. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    International Nuclear Information System (INIS)

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.)

  1. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-02-01

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.).

  2. RELAP5/MOD3 simulation for steam condensation under forced convection conditions

    International Nuclear Information System (INIS)

    Experimental and theoretical investigations were conducted by a team in the Department of Nuclear Engineering at the Massachusetts Institute of Technology (MIT) to determine the effects of noncondensable gases on steam condensation under forced convection conditions. The main objective of this study was to determine the condensation heat transfer coefficient of the steam in the presence of noncondensable gases, such as air and helium. In particular, the work was aimed at predicting the in-tube steam condensation rate as applied to the analysis of the isolation condensers of the proposed simplified boiling water reactor. The RELAP5 code uses laminar (Nusselt correlation) and turbulent film condensation (Carpenter ampersand Colburn correlation) heat transfer correlations in the absence of noncondensable gases, whichever is maximum. A reduction factor that is a function of the noncondensable gas concentration is being used to take into account the effect of the noncondensable gas on the condensation heat transfer coefficient. The properties for the gaseous phase are calculated assuming a Gibbs-Dalton mixture of steam and an ideal noncondensable gas. Since the experimental data are limited in the open literature, the MIT experimental program gives us an opportunity to assess the RELAP5 code against the separate-effects test data. The MIT test facility was simulated using the RELAP5 code for steam condensation in the presence of air under forced convection conditions. This paper presents RELAP5 simulation results of the MIT test facility for various inlet air mass fractions with fixed mixture inlet temperature by comparing with the MIT experimental data

  3. The RELAP5-Based NPA of the VVER Type Paks NPP

    International Nuclear Information System (INIS)

    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)

  4. Vertical downward subcooled bubbly flow modelling with RELAP5/MOD3.2.2 gamma

    International Nuclear Information System (INIS)

    The presented paper will consider the correlation for void fraction distribution in the subcooled boiling flow regime of downward liquid flow at low velocities. More specifically, it will focus on the choice of the most appropriate heat and mass transfer correlation. The experimental findings and theoretical consideration of these processes and phenomena will be compared with RELAP5/MOD3.2.2 Gamma predictions. (author)

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

    International Nuclear Information System (INIS)

    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

  6. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, N.; Khedr, A. [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt); D' Auria, F.D. [Pisa Univ. (Italy). Facolta di Ingegneria

    2015-12-15

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  7. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    International Nuclear Information System (INIS)

    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)

  8. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    International Nuclear Information System (INIS)

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  9. Application of the RELAP5 code for simulation of three turbine trip transients at the Peach Bottom Unit 2 BWR

    International Nuclear Information System (INIS)

    The RELAP5/MOD1, cycle 14 code was applied to simulation of three turbine trip tests conducted on the Peach Bottom Unit 2 boiling water reactor (BWR). The three turbine trip tests analyzed were initiated from different steady-state core power and primary flow conditions, thus providing a useful data set for evaluation of code performance and input modeling techniques. Simulation of these transients with RELAP5 required that the code be modified to include a jet pump model and to improve the RELAP5 steam separator model. The input model was derived from a RETRAN-01 model of the Peach Bottom Unit 2 Plant. Judgment was required in the selection of certain code input parameters such as moderator density and discharge coefficient. The final results obtained with the RELAP5 code were good simulations of all three Peach Bottom Unit 2 turbine trips for the first ten seconds of the transients

  10. Triangular coupling of SCDAP/Lower Head/RELAP5 using parallel virtual machine

    International Nuclear Information System (INIS)

    The present integrated SCDAP/RELAP5 MOD3.2 computer code is used for the simulation of reactor coolant system thermal-hydraulic (T/H) response and core damage progression. Due to the size of the code (∼150,000 lines), maintenance and upgrades present a significant burden. For the user, the code structure and data communication is difficult to understand, thus making it difficult to track down problems that arise during severe accident analyses. Adding to this difficulty is the fact that typical simulations can take anywhere from a few days to several weeks to complete. Therefore, in an effort to reduce the maintenance burden and enhance code performance, a change in the code structure was required. The U.S. Nuclear Regulatory Commission (USNRC) and the Swiss Federal Nuclear Safety Inspectorate (HSK), with technical assistance from SCIENTECH, Inc. and the Idaho Engineering and Environmental Laboratory (INEEL) initiated a concept of a triangular coupling of sub-modules within the SCDAP/RELAP5 MOD3.2 code: SCDAP, Lower Head (LOWHD), and RELAP5. The goal was to cleanly separate the SCDAP, LOWHD, and RELAP5 modules from the integrated SCDAP/RELAP5 code and then link these code modules using the Parallel Virtual Machine (PVM) software package. This triangular coupling using PVM is now successfully completed, and will facilitate the coupling of SCDAP/LOWHD with the new generation of thermal hydraulic codes, such as TRAC-M. With regard to parallel performance, this work is only a first step towards obtaining significant performance improvements. Further efforts will be required to fully realize these parallel performance gains. These are the further goals the USNRC and HSK propose to follow in the near future. The coupled code was verified by comparing results with those of the original merged code (i.e. the code as it existed prior to separating the three codes). First, three relatively simple developmental assessment cases were run, including a bundle boil

  11. An analysis of MB-2 100% steam line break test T-2013 using RELAP5/MOD2

    International Nuclear Information System (INIS)

    This report presents RELAP5/MOD2 calculations of the 100% steam line break test T-2013 performed on the Westinghouse Model Boiler-2 facility (MB-2). The input deck uses a noding structure typical of what would be used for an integral rig or full plant study using the RELAP5/MOD2 code. Sensitivity calculations were performed for the break junction discharge coefficient and the separator drain line loss coefficient. (author)

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

    International Nuclear Information System (INIS)

    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

  13. BLAZER: a RELAP5/MODI post processor to generate force-time history input data for structural computer codes

    International Nuclear Information System (INIS)

    A description is given of computer code BLAZER, a post processor to the RELAP5/MOD1 thermal hydraulics code. BLAZER computes fluid-induced piping forces and places them in structural code input format. The equations for properties and detailing the pressure, momentum, and acceleration components of the interaction force. A unique feature allows the number of time-force pairs describing the forcing functions to be reduced, since some structural codes such as NUPIPE-II limit this number. This reduction process maintains peaks and permits the concentration of force description points in time intervals where the forcing function experiences the most variation. The programs' plotting capability, which allows the analyst to compare the original and modified forcing functions and determine if proper reduction was achieved, is described. Input to structural codes SAP IV, NUPIPE-II, and ADINA can be produced by BLAZER. Capability to store intermediate results on tape is available. As an aid to the piping analyst, a step-by-step method for carrying out the calculation is included. 8 refs

  14. Pre-test of the KYLIN-II thermal-hydraulics mixed circulation LBE loop using RELAP5

    International Nuclear Information System (INIS)

    To investigate the behavior of lead bismuth eutectic (LBE) as coolant in China LEAd-based Research Reactor, Institute of Nuclear Energy Safety Institute (INEST), Chinese Academy of Sciences has built a multi-functional LBE experiment facility KYLIN-II. Mixed circulation loop, which is one of the KYLIN-II thermal-hydraulics loops, has the capability to drive the flowing LBE in different ways such as pump, gas lift and temperature difference (natural circulation). In this contribution, preliminary numerical simulations in support of the operation and experiment of KYLIN-II thermal-hydraulics mixed circulation LBE loop have been carried out and the obtained results have been studied. The RELAP5 Mod4.0 with LBE model has been utilized. Pre-test analysis showed the LBE circulation capability can reach the object under several driven patterns. The maximum velocity in fuel pin bundles can be larger than 0.15 m/s for natural circulation, 0.5 m/s for gas enhanced circulation, and 2 m/s for pump driven circulation. (author)

  15. Experiment and RELAP5 analysis for the downcomer boiling of APR1400 under LBLOCA

    International Nuclear Information System (INIS)

    Experiment has been carried out to investigate the boiling phenomena in the downcomer and RELAP5/MOD3.2 has been assessed with the present experimental data. The heated wall with a thickness of 8.2 cm and a height of 27 cm is used, which is made of the same material as the prototype (APR1400) with chrome coating against rusting. From the experiment, we visually observed strong liquid recirculation and vapor jetting near the heated wall due to the axial migration of voids only in the thin layer of the heated wall but little bubble migration to the bulk region. The size of the thin layer is about 4 cm, which is used for the determination of the radial nodal size in the radial double-node schemes. The RELAP5 calculations using three different nodal schemes are compared with experimental data in aspects of water level, void fraction, wall temperatures and phase velocities. The radial single-node scheme produces no liquid recirculation, resulting in the sudden level drop due to a sudden increase in void fraction. The double-nodal scheme with the top-bottom radial connections yields the strong circulation, eliminating the sudden level drop. As a result, the scheme produces better results than the radial single-nodal scheme and the double-nodal scheme with all radial connections. Based on the information from the measurement of local liquid velocity profile and visual observation, the drift velocity model is developed to apply into the downcomer with a large gap and a vertical heated wall. The proposed drift velocity model has been implemented into RELAP5 and verified with the experimental results. (author)

  16. Conversion of RELAP5 input data to PC [personal computer] spread sheet for steady-state hydraulic analysis

    International Nuclear Information System (INIS)

    An initial stage for any thermal-hydraulic system analysis is to derive the conditions that define steady-state operation. Often, complicated codes are utilized for this task; however, this represents unnecessary computational effort, since deriving steady-state performance is a relatively simple task. Incorporating a few correlations that describe specific behavior of a system into common personal computer spread-sheet software can be a useful tool in performing this task. One application of spread sheets is for generating steady-state hydraulic analysis of a system containing a single-phase fluid. Model developers can benefit from an additional source of information about the system and a medium for quick examination of potential model changes that this mechanism provides. This reduces the time the developer must spend and the computer usage needed to verify a models quality. The discussion presents a method and results for using information from RELAP5 to create a spread sheet of a full- or separate-effects system

  17. Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

  18. 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.

  19. Analysis of the natural circulation by the computer code RELAP-5

    International Nuclear Information System (INIS)

    The analysis of the natural circulation is one of the first analysis that was done at IJS with the computer code RELAP 5/MOS 1/CY 018. Specific model of the system was made for the natural circulation. The first 400 s of the transient were analyzed. At that time pumps are rotating only by coolant flow. First results show quite realistic picture of the transient although some changes should be made, especially on the steam generator model due to the unrealistic oscillations of the coolant flow on the secondary side. (author)

  20. Study on a data source for fault diagnosis of nuclear power plant based on RELAP5

    International Nuclear Information System (INIS)

    The model of nuclear power plant primary and secondary circuits was established, taking Qinshan Unit 1 as the object. The fault of SG U-tube rupture was calculated based on RELAP5 for solving the fault data source problem. Through analyzing the calculation results, we can know that simulation node is reasonable, input data card is exact, and data based on this model is credible. The designed data combining with fault diagnosis system has been debugged. The result indicates that the data is exact and enough and can be one of the databases for study on fault diagnosis of nuclear' power plant. (authors)

  1. RELAP5/MOD2 blind calculation of GERDA small break test and data comparison

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory (INEL), in support of the USNRC, has developed a RELAP5/MOD2 model of the GERDA facility to be used for analysis of the GERDA data, particularly relative to the phenomena of natural circulation and the boiler condenser mode of heat transfer. A blind calculation of GERDA Test 1605AA and a preliminary comparison with experimental data has been performed. The GERDA facility is a single loop integral facility with an electrically heated core. A general arrangement diagram of the facility is shown. The GERDA facility was designed for the performance of both separate effects and overall systems tests

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

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

    International Nuclear Information System (INIS)

    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

  4. Improvements to the RELAP5-3D Nearly-Implicit Numerical Scheme

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved with regard to its nearly-implicit numerical scheme for two phase flow and single-phase flow. Changes were made to the nearly-implicit numerical scheme finite difference momentum equations as follows: (1) added the velocity flip-flop mass/energy error mitigation logic, (2) added the modified Henry-Fauske choking model, (3) used the new time void fraction in the horizontal stratification force terms and gravity head, and (4) used an implicit form of the artificial viscosity. The code modifications allow the nearly-implicit numerical scheme to be more implicit and lead to enhanced numerical stability

  5. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  6. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

  7. Simulation of the LOFT L9-4 experiment with the code RELAP5/MOD2

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-02-01

    The integral effect test L9-4 (loss of off-site power without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transient in pressurized water reactors. The code was qualified for analysis of the thermal-hydraulics and kinetics associated to this kind of sequences. Some conclusions concerning simulation of anticipated transients without scram scenarios under natural circulation and axial power profile redistribution in power reactors are derived. (orig.).

  8. Simulation of the LOFT L9-4 experiment with the code RELAP5/MOD2

    International Nuclear Information System (INIS)

    The integral effect test L9-4 (loss of off-site power without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transient in pressurized water reactors. The code was qualified for analysis of the thermal-hydraulics and kinetics associated to this kind of sequences. Some conclusions concerning simulation of anticipated transients without scram scenarios under natural circulation and axial power profile redistribution in power reactors are derived. (orig.)

  9. Steady state and transient analyses of MNSR reactor using RELAP5 code

    International Nuclear Information System (INIS)

    Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.

  10. Steady state and transient analyses of MNSR reactor using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, E.; Khorsandi, Jamshid [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School; Tashakor, S. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School; Islamic Azad Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-03-15

    Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.

  11. Developmental assessment of the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    International Nuclear Information System (INIS)

    This report documents ten developmental assessment problems which were used to test the multidimensional component in RELAP5/MOD2.5, Version 3w. The problems chosen were a rigid body rotation problem, a pure radial symmetric flow problem, an r-θ symmetric flow problem, a fall problem, a rest problem, a basic one-dimensional flow test problem, a gravity wave problem, a tank draining problem, a flow through the center problem, and coverage analysis using PIXIE. The multidimensional code calculations are compared to analytical solutions and one-dimensional code calculations. The discussion section of each problem contains information relative to the code's ability to simulate these problems

  12. Improvements to the RELAP5-3D Nearly-Implicit Numerical Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Richard A. Riemke; Walter L. Weaver; RIchard R. Schultz

    2005-05-01

    The RELAP5-3D computer program has been improved with regard to its nearly-implicit numerical scheme for twophase flow and single-phase flow. Changes were made to the nearly-implicit numerical scheme finite difference momentum equations as follows: (1) added the velocity flip-flop mass/energy error mitigation logic, (2) added the modified Henry-Fauske choking model, (3) used the new time void fraction in the horizontal stratification force terms and gravity head, and (4) used an implicit form of the artificial viscosity. The code modifications allow the nearly-implicit numerical scheme to be more implicit and lead to enhanced numerical stability.

  13. 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.

  14. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    OpenAIRE

    2015-01-01

    Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed tha...

  15. Implementation of a new bubbly-slug interphase drag model in RELAP5/MOD2

    International Nuclear Information System (INIS)

    The implementation of a new bubbly-slug interphase drag model in the RELAP5/MOD2 code is described. The model is based on the determination of an effective interphase drag coefficient from a set of best-estimate void fraction correlations covering the full range of geometries and flow conditions encountered in PWR safety analysis. Calculations are reported which show that the new model leads to a much better prediction of void fraction profile for low flows in rod bundles than the standard model. Further work is necessary to derive a model formulation which can be guaranteed to produce physical drag coefficients in all flow situations. (author)

  16. Modeling with RELAP5/3.2. Thermal-hydraulic behaviour simulation because of the main pumps loss in the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Time evolution of natural circulation in the Atucha I nuclear power plant (CNA-I), in a main pumps lost incident because of the lost of external power feed, is analyzed. It leads to a strong stop transient, without an important blow down, from a forced nominal flow to a natural circulation one. The results are obtained from RELAP5/3.2 code's modeling. The study is based on the refrigeration condition analysis, during the first minutes of the reactor out of service. Previously to the transient, work had been done to obtain the plant steady state, with design parameters in operation conditions at 100 % of power. The object is that the actual plant state would be represented. In this way, each plant part (steam generators, reactor, pressurizer, pumps) had been modeled in separated form with the appropriate boundary conditions to be used in the whole circuit simulation. The developed model had been validated making use of the comparison between the values obtained to the principal thermodynamic parameters with the plant recorder values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/3.2 capability to model the thermodynamic phenomena in a PHWR plant has been verified when, according to the mentioned incident, the flow pass from a nominal forced flow, to one which is governed by natural circulation, still with the CNA-I untypical design conditions. (authors)

  17. Simulation of a small cold-leg-break experiment at the PMK-2 test facility using the RELAP5 and ATHLET codes

    International Nuclear Information System (INIS)

    Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of the experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation

  18. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  19. RELAP5/MOD3 modeling of water column rejoining and a water slug propelled by noncondensable gas

    International Nuclear Information System (INIS)

    The capability of the RELAP5/MOD3 computer code to analyze water hammer transients due to water column rejoining and a water slug propelled by noncondensable gas is investigated. The code-calculated results have been compared with those obtained from simple ideal analytical models. Good agreement is obtained between the calculation and analytical results in the initial period of the transient during which the water column or slug retains its sharp interface and suffers from little breakup or dissipation. As the transient proceeds, the code-calculated hydrodynamic loads are generally less than those implied by the analytical models. This is most likely due to the breakup of the water phase, which is not taken into account in the analytical models. Effects of time step and mesh sizes have also been studied. The results show that the usual Courant time limit applies. Finally, a sample calculation, corresponding to a water hammer transient in a typical Westinghouse four-loop reactor head vent system piping, is presented. The transient is induced by the opening of a relief valve and accelerating a trapped water slug through the pipeline. Hydrodynamic loads (i.e., force-time curves) on various pipe segments have been evaluated by appropriate postprocessing of the transient results. The calculated peak forces at selected pipe segments compare favorably with those estimated from the analytical models

  20. SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mladin, M. [Reactor Physics and Nuclear Safety Department, Institute for Nuclear Research-Pitesti, P.O. Box 78, Campului No. 1, 115400 Mioveni, Arges (Romania)], E-mail: mirea_mladin@easynet.ro; Dupleac, D.; Prisecaru, I. [Power Engineering Department, University ' Politehnica' of Bucharest (Romania)

    2009-02-15

    Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR-BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

  1. 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.

  2. Relap5 code analyses of parallel-tube behavior in horizontal heat exchanger of PCCS

    International Nuclear Information System (INIS)

    Parallel-tube behavior in a horizontal heat exchanger of a Passive Containment Cooling System (PCCS) for long-term cooling was analyzed using RELAP5/Mod 3.2 code. The heat exchanger composed of multiple U-tubes will be placed near the bottom of PCCS water pool. The length of heat exchanger tubes is different from tube to tube, according primarily to the radius of U-bend portion. A large temperature gradient and/or saturated boiling conditions would form in the PCCS secondary water pool as a result of natural convection in the course of the long-term cooling operation. The secondary water temperature and/or void fraction around the tube would vary along each tube. The change in the secondary cooling conditions would be more significant for the tube with a larger U-bend radius. The RELAP5/Mod3.2 code results indicated that the steam entrance flow rate would be insignificantly distributed although the tubes with different U-bend radius had large differences in the secondary cooling conditions. (author)

  3. Proposals for improving interphase drag modelling for the bubbly and slug regimes in RELAP5

    International Nuclear Information System (INIS)

    The proposal is put forward that the effective interphase drag coefficient for the bubbly and slug regimes in RELAP5 should be calculated using best-estimate void fraction correlations. It is argued that this will lead to improvements in the code's modelling of interphase drag and evidence is given to corroborate this. The need for such improvements has been prompted by the poor performance of the current models in simulating rod bundle experiments. There is also concern that the models do not account for profile slip effects, which could be important in a variety of geometries, and that the slug flow equations may not be appropriate for large diameter vertical pipes. To support the proposal, a set of void fraction correlations is identified which is believed to cover the full range of geometries and flow conditions encountered in PWR safety analysis including the analysis of small-scale experimental facilities. This set is selected from a detailed appraisal of the most appropriate correlations found in the literature which takes account of comparisons with experimental data and physical considerations. This Report forms part of the UK's commitment to the ICAP Code Improvement Plan. The recommendations will now be implemented in a development version of RELAP5/MOD3 and a preliminary assessment made. The interphase drag models used in the annular-mist regime will also be examined and, if necessary, appropriate improvements will be proposed. (author)

  4. Modeling Advanced Neutron Source reactor station blackout accident using RELAP5

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, and large-break LOCA results were presented and documented. This paper will focus on the station blackout scenario. The station blackout analyses have concentrated on thermal-hydraulic system response with and without accumulators. Five transient calculations were performed to characterize system performance using various numbers and sizes of accumulators at several key sites. The main findings will be discussed with recommendations for conceptual design considerations. ANS is a state-of-the-art research reactor to be built and operated at high heat flux, high mass flux, and high coolant subcooling. To accommodate these features, three ANS-specific changes were made in the RELAP5 code by adding: the Petukhov heat transfer correlation for single-phase forced convection in the thin coolant channel; the Gambill additive method with the Weatherhead wall superheat for the critical heat flux; and the Griffith drift flux model for the interfacial drag in the slug flow regime. 7 refs., 6 figs., 1 tab

  5. RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2010-01-01

    Full Text Available To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA. The current PSA standard recommends the use of best-estimate codes. The purpose of the study was to estimate the operator action success criteria time windows in scenarios in which the human actions are supplement to safety systems actuations, needed for updated human reliability analysis (HRA. For calculations the RELAP5/MOD3.3 best estimate thermal-hydraulic computer code and the qualified RELAP5 input model representing a two-loop pressurized water reactor, Westinghouse type, were used. The results of deterministic safety analysis were examined what is the latest time to perform the operator action and still satisfy the safety criteria. The results showed that uncertainty analysis of realistic calculation in general is not needed for human reliability analysis when additional time is available and/or the event is not significant contributor to the risk.

  6. RELAP5/MOD3.3 Code Validation with Plant Abnormal Event

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2008-07-01

    Full Text Available Measured plant data from various abnormal events are of great importance for code validation. The purpose of the study was to validate the RELAP5/MOD3.3 Patch 03 computer code with the abnormal event which occurred at Krško Nuclear Power Plant (NPP on April 10, 2005. The event analyzed was a malfunction, which occurred during a power reduction sequence when regular periodic testing of the turbine valves was performed. Unexpected turbine valve closing caused safety injection signal, followed by reactor trip. The RELAP5 input model delivered by Krško NPP was used. In short term, the calculation agrees very well with the plant measured data. In the long term, this is also true when operator actions and special plant systems are modeled. In the opposite, the transient would progress quite differently. Finally, the calculated data may be supplemental to plant measured data when the information is missing or the measurement is questionable.

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

    International Nuclear Information System (INIS)

    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

  8. Performance Assessment of Passive Heat Exchanger with Horizontal Tube using RELAP5

    International Nuclear Information System (INIS)

    In the world nuclear industry to enhance safety and reliability of nuclear power plant, passive safety system design has been introducing. Especially, passive auxiliary feedwater system (PAFS) has been applied to the advanced power reactor plus (APR+) in our domestic industry. According to PAFS design concept, PAFS makes role completely for the existing auxiliary feedwater system. PAFS can remove the residual heat in the core and then prevent the core damage when the feedwater is not available. The passive heat removal system has essentially heat exchanger with vertical or horizontal tubes. PAFS is a kind of passive heat exchanger with an inclined horizontal U tube bundle. Heat transfer phenomena in horizontal tubes play an important role in passive safety systems for the next generation of nuclear power plants. To assess the performance of the system, it is required to carry experiment and code analysis. NOKO experiment facility for investigating the emergency condenser effectiveness in SWR1000, is similar to PAFS. So the experiment result can be useful to evaluate the cooling performance of passive system like PAFS. The purpose of this study is to simulate the TH phenomena such as natural circulation and horizontal condensation heat transfer in NOKO experiment using RELAP5, and to compare the results between experimental data and RELAP5 code analysis

  9. Preliminary Performance Analysis on APR+ PAFS Using RELAP5 and MARS Codes

    International Nuclear Information System (INIS)

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with advanced technology. Domestic industry has been also developing a specific advanced reactor, so-called advanced power reactor plus (APR+), with passive auxiliary feedwater system (PAFS). The PAFS was introduced to replace an active auxiliary feedwater system (AFWS) completely. The system function is to remove the residual heat in the primary system like the AFWS does when the main feedwater system is unavailable. PAFS consists of a passive condensate cooling tank (PCCT), a heat exchanger, valves, and pipes as shown in Fig. 1. When PAFS works, steam from steam generator is supplied and condensed into water in the heat exchange. And the water falls down by gravity and returns to the steam generator. This progress goes on repeatedly and it makes natural circulation possible in the system. The circulating flow removes continuously the primary residual heat without any active components. The performance of PAFS depends on various thermo-hydraulic (TH) phenomena occurred in the system. So understanding these phenomena is required to analyze its performance with TH code such as RELAP5 or MARS which is a regulatory audit code. Licensee in the domestic industry has been conducting the performance analysis of the PAFS with RELAP5. So the analysis of MARS model with the same input was carried out and difference between two codes was compared in this study

  10. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

  11. A first look at LOCAs in the SBWR using RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Ghan, L.S.; Shaw, R.A.; Kullberg, C.M.

    1992-01-01

    The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE's Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.

  12. A first look at LOCAs in the SBWR using RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Ghan, L.S.; Shaw, R.A.; Kullberg, C.M.

    1992-12-31

    The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE`s Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.

  13. Design, experiments and Relap5 code calculations for the perseo facility

    International Nuclear Information System (INIS)

    Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories. The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe. ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained. An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses. This paper deals with the experimental and calculated results limited to the Relap5 code

  14. SCDAP/RELAP5 analysis of station blackout with pump seal LOCA in Surry plant

    International Nuclear Information System (INIS)

    During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB' analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Additionally, the calculated results were compared with the similar experimental studies of JAERI's ROSA-IV program. The present analyses showed that: (1) During S3-TMLB', the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB'. Accordingly, an intentional depressurization using PORVs is recommended for the mitigation of the accident consequences. (3) The present SCDAP/RELAP5 analyses did not show significant delay of accident progression. It was found that non-realistic lower heat generation and higher core cooling models used in the MELPROG/TRAC analysis are attributed to this discrepancy. (author)

  15. Atucha-2 obliquely inserted control rods RELAP5-3D/NESTLE model

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor in phase of commissioning in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g., oblique control rods, positive void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) model. In the framework of the agreement between NASA and University of Pisa a detailed NESTLE (three-dimensional neutron kinetics code) model of the Atucha-2 NPP was developed. This document summarizes the procedures for the implementation of the oblique control rods into the RELAP5/NESTLE model: a particular arrangement of RELAP5/NESTLE control rods insertion mode for such kind of oblique control rods and an implementation into the homogenized two group cross sections of ad-hoc calculated correction factors (these parameters were obtained by previously executed Monte Carlo calculations) was developed. Some applications, among the scenarios selected to perform safety analysis of the Atucha-2 NPP (CNA-II), are also reported: preliminary Scram Rod Worth, analysis of a Control Rod Ejection Accidents and a CR faulty withdrawal. (author)

  16. Thermalhydraulic analysis of Candu 6 100% reactor outlet header break using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I.; Ghitescu, P. [Power Plant Engineering Faculty, Politehnica University, Bucharest (Romania); Negut, G. [Institute for Nuclear Research, Pitesti (Romania)

    2007-07-01

    One of the postulated large break losses of coolant accident (LBLOCA) in a Candu reactor is the Reactor Outlet Header (ROH) break. After such an event, the normal coolant flow in the channels downstream the break is disrupted and late stagnation occurs after the pump head is degraded and before emergency core cooling (ECC) injection takes place. Given that the fuel decay and stored energy decreased significantly by that time, the heatup of the fuel sheath is much lower than in the case of 35% Reactor Inlet Header (RIH) break. However, the coolant pressure is much lower than the corresponding one at the 35% RIH break.The combination of a high fuel clad temperature and coolant low-pressure lead to more fuel failure events. Thus, the 100% ROH break has the highest potential for radioactivity release. The paper presents the thermal hydraulic analyses of a 100% reactor outlet header break. The study is done with RELAP5/SCDAP mod 3.4 and the results were compared with those of CATHENA. RELAP5 predicts a slightly faster inventory loss, an extended flow stagnation period and a higher clad temperature.

  17. 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.

  18. Preliminary Performance Analysis on APR+ PAFS Using RELAP5 and MARS Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Jin; Hong, Soon Joon [FNC Tech., SNU, Seoul (Korea, Republic of); Park, Ju Yeop; Seul, Kwang Won [KINS, Daejeon (Korea, Republic of)

    2010-10-15

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with advanced technology. Domestic industry has been also developing a specific advanced reactor, so-called advanced power reactor plus (APR+), with passive auxiliary feedwater system (PAFS). The PAFS was introduced to replace an active auxiliary feedwater system (AFWS) completely. The system function is to remove the residual heat in the primary system like the AFWS does when the main feedwater system is unavailable. PAFS consists of a passive condensate cooling tank (PCCT), a heat exchanger, valves, and pipes as shown in Fig. 1. When PAFS works, steam from steam generator is supplied and condensed into water in the heat exchange. And the water falls down by gravity and returns to the steam generator. This progress goes on repeatedly and it makes natural circulation possible in the system. The circulating flow removes continuously the primary residual heat without any active components. The performance of PAFS depends on various thermo-hydraulic (TH) phenomena occurred in the system. So understanding these phenomena is required to analyze its performance with TH code such as RELAP5 or MARS which is a regulatory audit code. Licensee in the domestic industry has been conducting the performance analysis of the PAFS with RELAP5. So the analysis of MARS model with the same input was carried out and difference between two codes was compared in this study

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

    International Nuclear Information System (INIS)

    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

  20. Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD3.1

    International Nuclear Information System (INIS)

    A small-break loss-of-coolant accident (SBLOCA) experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using RELAP5/MOD3.1. 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. The measured break mass flow rate was provided to participants as part of the initial conditions and is compared to that predicted by the RELAP5 model. Preliminary test results show that RELAP5 failed to predict the rise in heater rod temperature seen in the experimental data, possibly indicating a need to renodalize the core. Once all the post-test data is provided, a more detailed analysis of the predicted data will be performed. (author)

  1. Comparison between RELAP5/MOD1 and TRAC-PD2 computer codes in simulation of canon experience

    International Nuclear Information System (INIS)

    The present work reports comparisons between experimental and theoretical data done with the RELAP5/MOD1 and TRAC-FD2 codes, with particular emphasis on RELAP5/MOD1 code run with basic experimental data from the CANON depressurization simulation. This experiment simulates a Loss of Primary Coolant Accident due to a Large Rupture - LOCA, through the depressurization of a horizontal tube containing water with the instantaneous break of one side of the tube where measurements of pressure, temperature and void fraction are taken during the transient. The results of this comparison show that RELAP5/MOD1 code predict more satisfactorily the time dependent behavior of the pressure and void fraction than TRAC-PD2 code for several initial conditions considered in the CANON experiment. (Author)

  2. 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)

  3. Evaluation of the PASCAL experiment for quasi steady state and decrease of PCCT water level using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Jin; Hong, Soon Joon; Kim, Jeong Tae [SNU, Seoul (Korea, Republic of)

    2012-10-15

    In the field of nuclear engineering, there have been many researches to develop the passive safety system to enhance safety and reliability of nuclear power plant. In South Korea, the development of PAFS (Passive Auxiliary Feedwater System) is ongoing to be applied to the advanced power reactor plus (APR+). It can replace completely a conventional auxiliary feedwater system. KAERI (Korea Atomic Energy Research Institute) constructed the test facility named PASCAL (PAFS Condensing Heat Removal Assessment Loop) to validate the cooling and operational performance of the PAFS. Its dimension and material are same as the proto type U tube of the PAFS. In this study, the RELAP5 calculations for PASCAL experiment were performed to evaluate its cooling performance under quasi steady state and long term cooling condition. From the comparison between the RELAP5 analysis results and PASCAL experiment, the improvement of the RELAP5 input model for PASCAL was performed.

  4. Evaluation of the PASCAL experiment for quasi steady state and decrease of PCCT water level using RELAP5

    International Nuclear Information System (INIS)

    In the field of nuclear engineering, there have been many researches to develop the passive safety system to enhance safety and reliability of nuclear power plant. In South Korea, the development of PAFS (Passive Auxiliary Feedwater System) is ongoing to be applied to the advanced power reactor plus (APR+). It can replace completely a conventional auxiliary feedwater system. KAERI (Korea Atomic Energy Research Institute) constructed the test facility named PASCAL (PAFS Condensing Heat Removal Assessment Loop) to validate the cooling and operational performance of the PAFS. Its dimension and material are same as the proto type U tube of the PAFS. In this study, the RELAP5 calculations for PASCAL experiment were performed to evaluate its cooling performance under quasi steady state and long term cooling condition. From the comparison between the RELAP5 analysis results and PASCAL experiment, the improvement of the RELAP5 input model for PASCAL was performed

  5. 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)

    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

  6. Assessment of interfacial shear and wall heat transfer of RELAP5/MOD2/36.02 during reflooding

    International Nuclear Information System (INIS)

    The analysis of a number of reflooding and one boil-off experiment in the electrically heated rod bundle NEPTUN at EIR with RELAP5/MOD2/36.02 showed significant differences between measurements and predictions. The same was true for the analysis of two FLECHT-SEASET reflooding experiments. In this work, we shall report on these items and we shall present the modifications made to the frozen version of RELAP5/MOD2/36.02 which eliminate most of the observed discrepancies. (author)

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

    International Nuclear Information System (INIS)

    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.

  8. RELAP5-3D Developmental Assessment: Comparison of Version 4.2.1i on Linux and Windows

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

    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.

  10. PUMA-PCCS separate effect tests and RELAP5 code evaluation in PUMA

    Science.gov (United States)

    Choi, Sung Won

    One of the key areas in the design of advanced nuclear reactors is to develop a reliable Passive Containment Cooling System (PCCS). The purpose of the current work is to better understand the condensation phenomena in PCCS for the downward co-current flow of a steam/air mixture through condenser tube bundles during the three PCCS operational modes, namely the bypass mode, the cyclic venting mode and the long-term cooling mode. A series of unique separate-effect PCCS test data were obtained for condensation heat transfer in the PCCS heat exchangers of the PUMA (Purdue University Multidimensional Integral Test Assembly) facility under a task sponsored by the U.S. Nuclear Regulatory Commission. Test conditions includes bypass mode, cyclic venting mode and long term mode, covering a wide range of Loss of Coolant Accident(LOCA) conditions with a parameters of pressure, mass flow rate, noncondensable(NC) gases, and PCCS pool water level. The parametric effect studies and a further validation of the PUMA-PCCS separate effect test data were performed. The evaluation of a best estimate system code (RELAP5/MOD3.3) was performed by using unique PUMA-PCCS separate effects data and PUMA-Main Steam Line Break (MSLB) integral test (1998). Through a sensitivity studies of nodalization method and physical models on the MSLB test simulations, deficiencies in RELAP5/MOD3.3 code were found as follows: (1) over prediction of heat removal rate by condensation models, (2) overestimation of SP heat transfer through the horizontal venting line and thermal stratification distortion, (3) underestimation of NC gas effects in PCCS by the distortion of cyclic venting phenomena and (4) overestimation of the DW and SP wall condensation. The improvement for the code calculation predictions could be obtained by removing the RELAP5/MOD3.3 code deficient factors in the PUMA MSLB integral test simulation. The unique PCCS NC gas venting visualizations were obtained according to various PCCS inlet NC

  11. Assessment of conservatism embedded in licensing calculations of loss of coolant accident via RELAP5-3D/K simulation of LOFT L2-3 experiment

    International Nuclear Information System (INIS)

    Experiment L2-3 is one of loss-of-coolant experiment (LOCE) in Loss of Fluid Test (LOFT) program conducted by Idaho National Engineering and Environmental Laboratory (INEEL). The experiment mimics the double-ended cold leg large break Loss of Coolant Accident (LOCA) of a commercial Pressurized Water Reactor (PWR). The primary objectives of experiment are to determine fuel rod cladding thermal response, determine the performance of the Emergency Core Cooling System (ECCS) and determine the core reflood characteristics during a LOCA. In the present study, 1-D version of RELAP5-3D and RELAP5-3DK codes are used to simulate L2-3 LOCE. RELAP5-3D is a best estimated system thermal hydraulic analysis code of commercial nuclear power reactors. RELAP5-3D/K is a modified version of RELAP5 code series, which embedded RELAP5-3D with evaluation models required in U.S. Code of Federal Regulations 10 CFR 50 Appendix K for LOCA licensing calculation. The simulation and experiment results are compared with one another to assess the conservatism embedded in RELAP5-3D/K code. The result shows that both RELAPD5-3D and RELAP5-3D/K have reasonable system responses in simulation of L2-3 test. RELAP5-3D result shows cladding temperature behavior similar to experiment result at hot spot. RELAP5-3D/K demonstrates conservatism in LOCA calculation. The peak cladding temperature is 60 K higher than experiment result. Sensitivity studies show that the way that the crossflow junctions in the core are connected has significant impact on the predicted peak cladding temperature of RELAP5-3D results. (author)

  12. Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Cliff B. Davis

    2010-09-01

    RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

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

    International Nuclear Information System (INIS)

    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 MWt 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-08

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

  15. Uncertainty study on RELAP5 reflood post-CHF heat transfer model

    International Nuclear Information System (INIS)

    In the best estimate plus uncertainty analysis of large break loss of coolant accident (LBLOCA), the key is the study on uncertainty analysis of reflood post-CHF heat transfer model. In this paper, the uncertainty of RELAP5 reflood post-CHF heat transfer model was studied. The transition boiling model of reflood was evaluated through the transition boiling test data at 0.1-0.4 MPa by Weisman and the film boiling model of reflood was evaluated through the film boiling test data at low pressure by Idaho National Engineering Laboratory (INEL). The probability distribution function of the model was obtained, which can be used in the LBLOCA uncertainty analysis. (authors)

  16. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    International Nuclear Information System (INIS)

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  17. Vectrorization of the LWR transient analysis code RELAP5/MOD2/CYCLE36

    International Nuclear Information System (INIS)

    The LWR transient analysis code RELAP5/MOD2/CYCLE36 has been vectorized. The performance of the vectorized code in vector mode of the FACOM VP-100 was 3.5 times higher than that of original one in scalar mode for a real scale transient calculation. Subroutines for heat conduction were vectorized in terms of heat structures and heat meshes, while those for hydrodynamics were vectorized in terms of volumes and junctions. In this report, the vectorization method used for each of subroutines and its vectorization effect are described. A reduction of the overhead of bit operation functions which is introduced when the code was converted from the CDC to the FACOM is also described. (author)

  18. Analysis Of Control Rod Ejection Of APR1400 By RELAP5

    International Nuclear Information System (INIS)

    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)

  19. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    Directory of Open Access Journals (Sweden)

    Patrícia A. L. Reis

    2015-01-01

    Full Text Available Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

  20. Interpretation of TRIGA reactivity transients with RELAP5/PARCS coupled-code

    International Nuclear Information System (INIS)

    In the frame of future experiments to carried out upon TRIGA reactors, which aim to verify the real feasibility of the ADS (Accelerator Driven System) concept, it is essential to build a numerical tool able to simulate the dynamic behaviour of the reactor in subcritical configuration. This model developed to support the design of subcritical experiments and the safety analysis of the reactor, as a first step has to be assessed against the experimental data available for the critical reactor. To this purpose the thermal-hydraulic/ neutronic numerical model based on the RELAP5/PARCS coupled-code is been tested against the experimental reactivity transients conducted on the RC1-TRIGA reactor at the ENEA Casaccia Research Center in forecast of the TRADE (TRIGA Accelerator Driven Experiment) subcritical experience. The results of the calculations already performed show a qualitative good agreement with the experimental data and allow to address the future developments and improvements of the numerical model. (authors)

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

    International Nuclear Information System (INIS)

    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)

  2. Thermal hydraulics safety analysis of Candu reactor using the RELAP5 system code

    International Nuclear Information System (INIS)

    In this study, the response of CANDU-6 nuclear reactor to several transients are investigated. The simulation of the system is performed by using RELAP5 thermalhydraulic system code. AECL performes the transient simulations of CANDU reactor by using the FIREBIRD code, developed by AECL for thermal hydraulic analysis of CANDU. All analysis for LOCA and ECCS effectiveness were done by using the FIREBIRD code. The investigations concerning the RELAP5 analysis of CANDU system are too few. Better normal operating conditions are achieved, the effect of pipe interconnecting the outlet headers in a loop is observed. It is found that, with the reactor outlet headers interconnected, the system is stable to perturbations but would exhibit divergent pressure, quality and flow oscillations if the interconnection is removed and if the quality at the reactor outlet header region is greater than 1-2% but less than 8%, specific large (100% of flow area) and small (10% of flow area) breaks in both inlet and outlet headers and in the pump suction are analysed. Results indicate that, l00% break in the inlet header has more probability of fuel failure than the same size break in the outlet header. The worst break location is found to be the pump suction with a break size of 100%. Higher void fractions, higher outlet header quality and heat temperatures are observed in the large break transients than that of small break transients. For small break transients, the break location in the inlet header results higher void fraction, outlet header quality and sheath temperatures than that of outlet header break transients. Emergency core cooling system (ECCS) is found to be effective for the cases analysed. Initiating trip parameters and time for scram and ECCS injection is also investigated

  3. NPP Krsko RELAP5 Analysis of the Scaled Loss of RHR Experiment Scenario

    International Nuclear Information System (INIS)

    In the paper the RELAP5/mod322 analysis of the BETHSY test 6.9 c scaled to NPP Krsko is presented. The BETHSY test 6.9 c consisted in a loss of the Residual Heat Removal (RHR) system during mid-loop operation and with pressurizer and SG 1 manway open. Term mid-loop designates the condition with reduced water level close to the center elevation of the horizontal legs in order to enable maintenance and inspection activities during the plant outage. Under low pressure conditions loss of the RHR system causes a quick boiling with subsequent liquid disposal and spillage of the primary system inventory through the manway openings. Transient outcome (time to start of boiling, core liquid level) depends on core decay level, initial liquid inventory as well as of the size and position of the openings in the system. At the beginning of the BETHSY test, the liquid was at saturation temperature and the rest of the facility was filled with steam at atmospheric pressure. Aims of the BETHSY test 6.9 c were to identify and characterize the physical phenomena encountered under conditions of very low power and low pressure accompanied by inventory loss through the openings. A set of sensitivity study calculations of the BETHSY test 6.9 c scaled to NPP Krsko was performed. Aims of the sensitivity calculations and of the comparison with experimental data are manifold. First, the appropriateness of the developed RELAP5/mod322 NPP Krsko nodalization and the particular transient consequences can be assessed. Secondly, the assessment of the influences of different scaling approaches and initial conditions help better characterize physical phenomena during mid-loop operation and choosing proper scaling as well as transient modeling approach. (author)

  4. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  5. 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.

  6. Theory and input requirements for the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    International Nuclear Information System (INIS)

    This report documents the theory and input requirements for the multidimensional component in RELAP5/MOD2.5, Version 3w. The equations in Cartesian and cylindrical coordinates are presented as well as the shallow water terms. The implementation of these equations is then discussed. Finally, the constitutive models and input requirements are then described

  7. Loss of main pumps in the Atucha 1 nuclear power station. Modeling with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    This paper analyzes the temporal evolution of natural circulation at the Atucha 1 nuclear power station when losing its main pumps due to lack of external feeding power. This leads to a temporal stoppage with important depressurization, from the nominal flow of forced circulation to another of natural one. Results are attained with a program called RELAP5/MOD3.2

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

  10. Modification of blowdown heat transfer models for RELAP5-3D in accordance with appendix K of 10CFR50

    International Nuclear Information System (INIS)

    The objective of this paper is to implement the blowdown heat transfer models accepted by Appendix K of 10CFR50 into RELAP5-3D and to rename it as RELAP5-3D/K. Modifications of critical heat flux (CHF) model, post-CHF model, and the heat transfer logic for nucleate and transition boiling lockout are included. Also the assessments against separate-effect experiments were evaluated for RELAP 5-3D/K. From calculation results, the conservative predictions of surface peak temperatures using RELAP5-3D/K are obtained. It demonstrated that the blowdown heat transfer models were successfully modified and implemented into RELAP5-3D in accordance with Appendix K of 10CFR50. (authors)

  11. Modification of blowdown heat transfer models for RELAP5-3D in accordance with appendix K of 10CFR50

    Energy Technology Data Exchange (ETDEWEB)

    Chin-Jang, Chang; Liang, T.K.S. [Nuclear Engineering Div. Institute of Nuclear Energy Research, Lung-Tan, Taiwan (China); Huan-Jen, Hung; Wang, L.C. [Power Research Institute, Taiwan Power Company (China)

    2001-07-01

    The objective of this paper is to implement the blowdown heat transfer models accepted by Appendix K of 10CFR50 into RELAP5-3D and to rename it as RELAP5-3D/K. Modifications of critical heat flux (CHF) model, post-CHF model, and the heat transfer logic for nucleate and transition boiling lockout are included. Also the assessments against separate-effect experiments were evaluated for RELAP 5-3D/K. From calculation results, the conservative predictions of surface peak temperatures using RELAP5-3D/K are obtained. It demonstrated that the blowdown heat transfer models were successfully modified and implemented into RELAP5-3D in accordance with Appendix K of 10CFR50. (authors)

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

    International Nuclear Information System (INIS)

    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)

  13. 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)

  14. SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

    International Nuclear Information System (INIS)

    Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles, and the steam generator (SG) tubes. The potential for a steam generator tube rupture (SGTR) is of particular concern because fission products could be released to the environment through such a failure. The Nuclear Regulatory Commission (NRC) developed a program to address SG tube integrity issues in operating pressurized water reactors (PWRs) based on the possibility for environmental release. An extensive effort to evaluate the potential for accident-induced SGTRs using SCDAP/RELAP5 at the Idaho National Engineering and Environmental Laboratory (INEEL) was directed as one part of the NRC program. All SCDAP/RELAP5 calculations performed during the INEEL evaluation were based on station blackout accidents (and variations thereof) because those accidents are considered to be one of the more likely scenarios leading to natural circulation flows at temperatures and pressures that could threaten SG tube integrity (as well as the integrity of other vulnerable RCS pressure boundaries). Variations that were addressed included consideration of the effects of RCP seal leaks, intentional RCS depressurization through pressurizer PORVs, SG secondary depressurization, DC-HL bypass flows, U-tube SG sludge accumulation, and quenching of upper plenum stainless steel upon relocation to the lower head. Where available, experimental data was used to guide simulation of natural circulation flows. Independent reviews of the applicability of the natural circulation experimental data, the suitability of the code, and the adequacy of the modeling were completed and review recommendations were incorporated into the evaluation within budget and

  15. Verification of RELAP5/MOD3 with theoretical and numerical stability results on single-phase, natural circulation in a simple loop

    International Nuclear Information System (INIS)

    The theoretical results given by Pierre Welander are used to test the capability of the RELAP5 series of codes to predict instabilities in single-phase flow. These results are related to the natural circulation in a loop formed by two parallel adiabatic tubes with a point heat sink at the top and a point heat source at the bottom. A stability curve may be defined for laminar flow and was extended to consider turbulent flow. By a suitable selection of the ratio of the total buoyancy force in the loop to the friction resistance, the flow may show instabilities. The solution was useful to test two basic numerical properties of the RELAP5 code, namely: a) convergence to steady state flow-rate using a 'lumped parameter' approximation to both the heat source and sink and; b) the effect of nodalization to numerically damp the instabilities. It was shown that, using a single volume to lump the heat source and sink, it was not possible to reach convergence to steady state flow rate when the heated (cooled) length was diminished and the heat transfer coefficient increased to keep constant the total heat transferred to (and removed from) the fluid. An algebraic justification of these results is presented, showing that it is a limitation inherent to the numerical scheme adopted. Concerning the effect of nodalization on the damping of instabilities, it was shown that a 'reasonably fine' discretization led, as expected, to the damping of the solution. However, the search for convergence of numerical and theoretical results was successful, showing the expected nearly chaotic behavior. This search lead to very refined nodalization. The results obtained have also been verified by the use of simple, ad hoc codes. A procedure to assess the effects of nodalization on the prediction of instabilities threshold is outlined in this report. It is based on the experience gained with aforementioned simpler codes. (author)

  16. Assessment of SCDAP/RELAP5 modelling of top-down reflood, using PARAMETER bundle experiments

    International Nuclear Information System (INIS)

    The PARAMETER experimental programme was conducted by LUCH under the oversight of the International Science and Technology Centre (Russia), to investigate core heatup and reflooding. The main focus was on reflood by top or combined injection, under beyond Design Basis Accident (DBA) conditions but with intact geometry. The test bundle was electrically heated but the materials and configuration were otherwise prototypic of a VVER, that is Zr-1%Nb (E-110) cladding, hexagonal rod geometry and unirradiated UO2 for the pellet simulation. These attributes make the PARAMETER test data unique. A limited set of post-test calculations of the combined injection experiment SF2 were performed using a local version of SCDAP/RELAP5 which includes a model for the tantalum electrical heater elements used in PARAMETER, to support development and benchmarking of the input deck to be used for later tests. One of the aims of this analysis was to identify a suitable oxidation correlation for the E- 110 cladding. the A specific feature of SF3 was the top injection, initiated at a maximum temperature of ca. 1815 K. Pre-test calculations were performed for SF3 to support the test definition and protocol, and to provide a starting point for post-test analyses using the same models. Despite some discrepancies between post-test calculation and data during the final heatup prior to reflood initiation, it was possible to specify the conditions so that the calculated temperatures at the start of injection were sufficiently in agreement to provide a basis for assessment of the reflood modelling. The main characteristic of the observed reflood and quench progression, a slow top-down quench to about half way followed by a more rapid bottom up quench, was not reproduced when the standard RELAP5 model for interphase friction was used. However, use of a Wallis-type counter-current flow limitation (CCFL) model, derived from recent tests with representative upper tie-plate geometry, successfully

  17. A Study on the Flow Capacity of PAFS using RELAP5/MOD3.3

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Seong Su; Lee, Kyung Jin; Hong, Soon Joon [FNC Tech., SNU, Seoul (Korea, Republic of); Kang, Sang Hee; Cheon, Jong; Kim, Han Gon [NETEC, Daejeon (Korea, Republic of)

    2010-10-15

    In South Korea, advanced power reactor plus (APR+), as a Korean specific reactor, is currently under development for the export strategy. In order to raise competitiveness of the APR+ in the world market, it is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a passive auxiliary feedwater system (PAFS) was adopted as an improved safety design concept of APR+: and then there have been many efforts to develop the PAFS. The design concept of PAFS is as follows 1): 1) PAFS can completely replace the auxiliary feedwater system (AFWS). 2) When the design basis accident (DBA), in which feedwater is not available, occurs, the PAFS can remove the residual heat in the core and then prevent the core damage. 3) PAFS is operated by the natural circulation of the condensed steam due to the condensation and gravity force: and then reduces the operator action for the reactor safety. In order to confirm whether the PAFS can actually replace the AFWS in various accidents, it is required to carry out the performance analysis of the PAFS. For that, the main purposes of this study are: 1) to develop the RELAP5 input model, 2) to analyze the performance of the PAFS after applying the PAFS model into the RCS input model, 3) to produce the AFW flowrate to be used in the design safety analysis code. Up to earlier this year, 2010, PAFS RELAP5 input model has been developed/improved by using the newly updated design data; nowadays, the performance analyses for various accidents such as loss of condenser vacuum (LOCV), feed line break (FLB), and steam generator tube rupture (SGTR) is ongoing with APR1400 input model. Moreover, in order to produce the minimum AFW flowrate to be used in design safety analysis code, new methodology was developed and tested. This paper focuses on the new methodology for the production of the minimum AFW flowrate and the sensitivity analyses for the heat source/sink

  18. A Study on the Flow Capacity of PAFS using RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    In South Korea, advanced power reactor plus (APR+), as a Korean specific reactor, is currently under development for the export strategy. In order to raise competitiveness of the APR+ in the world market, it is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a passive auxiliary feedwater system (PAFS) was adopted as an improved safety design concept of APR+: and then there have been many efforts to develop the PAFS. The design concept of PAFS is as follows 1): 1) PAFS can completely replace the auxiliary feedwater system (AFWS). 2) When the design basis accident (DBA), in which feedwater is not available, occurs, the PAFS can remove the residual heat in the core and then prevent the core damage. 3) PAFS is operated by the natural circulation of the condensed steam due to the condensation and gravity force: and then reduces the operator action for the reactor safety. In order to confirm whether the PAFS can actually replace the AFWS in various accidents, it is required to carry out the performance analysis of the PAFS. For that, the main purposes of this study are: 1) to develop the RELAP5 input model, 2) to analyze the performance of the PAFS after applying the PAFS model into the RCS input model, 3) to produce the AFW flowrate to be used in the design safety analysis code. Up to earlier this year, 2010, PAFS RELAP5 input model has been developed/improved by using the newly updated design data; nowadays, the performance analyses for various accidents such as loss of condenser vacuum (LOCV), feed line break (FLB), and steam generator tube rupture (SGTR) is ongoing with APR1400 input model. Moreover, in order to produce the minimum AFW flowrate to be used in design safety analysis code, new methodology was developed and tested. This paper focuses on the new methodology for the production of the minimum AFW flowrate and the sensitivity analyses for the heat source/sink

  19. Analysis of void fraction in single channel using TRACE, MARS-KS, and RELAP5

    International Nuclear Information System (INIS)

    The accurate prediction of the void fraction is one of the most important factors in subchannel analyses. In general, the subchannel analysis has been conducted by means of dedicated subchannel analysis codes. However, the recent development of system codes with advanced two-phase flow model and three-dimensional components has helped more accurate and precise prediction of multiphase phenomena in nuclear reactor systems. This paper aims at evaluating the applicability of three different system codes to the prediction of the void fraction based on NUPEC experiment results employed for OECD/NRC PSBT benchmark. As a first step to the full scope analysis, analysis results for single channel experiments by using TRACE 5.0 Patch 3, MARS-KS 1.3, and RELAP5/MOD3.3 Patch 4 are presented in this paper. The result indicates that all codes slightly over-predict the void fraction compared to the experimental results in general and no significant discrepancies between the codes are observed. (author)

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

    International Nuclear Information System (INIS)

    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

  1. Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

    International Nuclear Information System (INIS)

    Highlights: ► We investigated how much the code results are affected by the code user. ► Two essential modifications were made on a previously validated nodalization. ► We used the RELAP5 code to predict the results. ► Results highlight the necessity of sensitivity analysis to have the ideal modeling. - Abstract: The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.

  2. RELAP5/MOD2 shutdown thermal-hydraulic analysis for NPP Krsko

    International Nuclear Information System (INIS)

    The purpose of the previous simplified and very conservative shutdown thermalhydraulic analysis [2] was to determine the ''time to boiling'' and ''time to core uncover'' in the event of a loss of Decay Heat Removal (DHR). The various Reactor Coolant System (RCS) levels and pressure boundaries to predict necessary containment closure time (CCT) were assumed. RCS heatup rates, the required makeup rates (required flow into RCS to prevent boiling) and required RCS mass flows to maintain subcooling were derived. Main purpose of that analysis was to provide input to overall analysis of plant safety during shutdown. The previous analysis [2] was performed with the most conservative assumptions like thermal isolation of reactor coolant loops from core, etc. The loss of water through pressurizer during a loss of DHR on midle loop operation was not considered. This paper presents re-analysis of chosen case (Pressurizer Manway Open - PMO) by a simplified analysis with best-estimate RELAP5/MOD3.2.2 gamma calculation taking into account reference initial and boundary conditions and modeled geometrical correspondence between plant zones and control volumes. There is also investigation of possible discharging flow via PMO is considered. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  5. A web-based nuclear simulator using RELAP5 and LabVIEW

    International Nuclear Information System (INIS)

    A web-based nuclear reactor simulator has been developed using the best-estimate nuclear system analysis code RELAP5 as its engine, and LabVIEW for graphical user interface and web-casting. Simulator retains the accuracy of the best-estimate code. Results are displayed in user friendly graphical format. Color-coded nominal values are displayed along with the current status of different variables in tab activated windows. Some variables of interest are also shown as a function of time. All graphical outputs are displayed in web browsers making the simulator's front end independent of the operating system. The interactive simulation feature allows the users to simulate specific reactor transients - such as LOCA, scram, etc. - using a single click. Simulator's graphical output can be web-casted and is thus available to anybody with access to the web. Moreover, if permitted, the simulator can be operated remotely from another site connected to the server via the World Wide Web

  6. Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

    2011-07-01

    The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

  7. Station blackout accidents for the Korea Nuclear Unit 1 using RELAP5/MOD1

    International Nuclear Information System (INIS)

    A station blackout accident which occured at the Korea Nuclear Unit 1 (KNU-1) at the Kori site in Korea on June 9, 1981 was analyzed by using the RELAP5/MOD1 code. The incident was occured at 11:05 a.m. due to the malfunction of a steam generator level gauge. The false level signal eventually caused the reactor and turbine trip. Following the turbine trip, the excitor of the generator remained functioning and the reactor coolant pumps remained connected to the internal source for 30 seconds, thus providing full reactor coolant flow for 30 seconds after the reactor trip. Upon the loss of the generator power, one of two buses failed to automatically transfer to the off-site power and the other also failed in 30 seconds after generator trip. The transfer to the off-site power was restored in about 26 minutes. During the blackout period two diesel generators provided the necessary electrical power to the corresponding instruments and two motor-driven auxiliary feedwater pumps

  8. Uncertainty and sensitivity analyses of the Kozloduy pump trip test by coupled RELAP5/PARCS code

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia Salah, A. [Pisa Univ., Facolta di Ingegneria, DIMNP (Italy); Kliem, S.; Rohde, U. [Forschungszentrum Rossendorf (FZR) (Germany)

    2005-07-01

    The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulating and more precise consideration of multidimensional effects under complex transients in NPPs. In the current paper a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data and the comparison study shows good agreements between the calculations and the global kinetic and thermal-hydraulic aspects observed experimentally. It appears that at steady state level, the simulation errors are mainly due to the absence of ADF correction and to the model used for evaluating the Doppler feedback effect. During the transient, the discrepancies are mainly due to the combined effect of uncertain parameters related to the measurement of control rod course, and the estimation of the Doppler effect.

  9. Reformulation RELAP5-3D in FORTRAN 95 and Results

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2010-08-01

    RELAP5-3D is a nuclear power plant code used worldwide for safety analysis, design, and operator training. In keeping with ongoing developments in the computing industry, we have re-architected the code in the FORTRAN 95 language, the current, fully-available, FORTRAN language. These changes include a complete reworking of the database and conversion of the source code to take advantage of new constructs. The improvements and impacts to the code are manifold. It is a completely machine-independent code that produces machine independent fluid property and plot files and expands to the exact size needed to accommodate the user’s input. Runtime is generally better for larger input models. Other impacts of code conversion are improved code readability, reduced maintenance and development time, increased adaptability to new computing platforms, and increased code longevity. The conversion methodology, code improvements and testing upgrades are presented in a manner that will be useful to future conversion projects for other such large codes. Comparison between the pre- and post-conversion code are made on the basis of code metrics and code performance.

  10. Independent assessment of TRAC and RELAP5 codes through separate effects tests

    International Nuclear Information System (INIS)

    Independent assessment of TRAC-PF1 (Version 7.0), TRAC-BD1 (Version 12.0) and RELAP5/MOD1 (Cycle 14) that was initiated at BNL in FY 1982, has been completed in FY 1983. As in the previous years, emphasis at Brookhaven has been in simulating various separate-effects tests with these advanced codes and identifying the areas where further thermal-hydraulic modeling improvements are needed. The following six catetories of tests were simulated with the above codes: (1) critical flow tests (Moby-Dick nitrogen-water, BNL flashing flow, Marviken Test 24); (2) Counter-Current Flow Limiting (CCFL) tests (University of Houston, Dartmouth College single and parallel tube test); (3) level swell tests (G.E. large vessel test); (4) steam generator tests (B and W 19-tube model S.G. tests, FLECHT-SEASET U-tube S.G. tests); (5) natural circulation tests (FRIGG loop tests); and (6) post-CHF tests (Oak Ridge steady-state test)

  11. Total loss of CNA1 steam generators feed water simulated with RELAP5/MOD3

    International Nuclear Information System (INIS)

    The results of the calculations are presented carried out by utilizing the code RELAP5/MOD3, upon the basis of the postulated initial event of total loss of feed water to the two steam generators in the nuclear power plant Atucha 1, CNA1. The evolution of the installation systems during the transient was analyzed in different conditions of availability: condenser, relief valve and safety valves in the secondary system, safety valves in the primary system and system of long-term subsequent cooling. Located in the primary and secondary systems of the installation they turn out to be prominent in this event. Upon this basis the sequences of possible evolution were calculated and those that would conduct the system toward the setting called 'damage to the core' were determined. Also those in which would arrive to a state of 'safe shutdown' were determined. These results were utilized in the verification of the tree of events utilized in the Final Report of the Probabilistic Safety Analysis for the sequence of event T9, made from calculations carried out with the code DINETZ. From this compare some differences were determined and are presented in the modified version of tree of events. (author)

  12. Numerical experiments in concurrent multiprocessing with the RELAP5 nuclear reactor systems code

    International Nuclear Information System (INIS)

    Numerical experiments performed on a single instruction multiple data-pipeline vector parallel (SIMD-PVP) architecture computing machine, e.g., a CRAY X-MP/48, demonstrate that current nuclear reactor systems codes can be restructured for concurrent multiprocessing and show wall clock performance improvements of 1.5 to 3.0 on a 4-CPU machine, depending on plant model, problem type, and problem length. In addition, algorithm development studies indicate that up to a 20% speedup can be obtained by a new class of parallel numerical methods. Faster-than-real-time simulation has been demonstrated utilizing RELAP5/MOD1 and a pressurized water reactor plant model characteristic of licensing and/or safety analysis calculations. A theoretical analysis indicates that five to ten times faster than real-time computation may be possible for this class of problems utilizing this or the next generation of SIMD-PVP architecture machines, such as the CRAY X-MP/48, and new computer codes optimized for such machines

  13. Preliminary use of RELAP5 Code with Internal Assessment of Uncertainty

    International Nuclear Information System (INIS)

    The present work deals with the preliminary use of the RELAP5 code with the internal assessment of the uncertainty. The activity has been started as a follow up of the research leading to the proposal and applicability of UMAE (Uncertainty Methodology base on Accuracy Extrapolation) uncertainty methodology. The concept of hypercube to characterize the status of a LWR plant during any transient and the assumption of uncertainty connected with the hypercubes are adopted: the CIAU (Code with Internal Assessment of Uncertainty) method has been set up. In the frame of the activity of CIAU development, matrix and a vector for uncertainty have been obtained; three type of matrixes and vectors are utilized: 1. matrix and vector deriving from tests results qualified by UMAE conditions; 2. matrix and vector deriving from tests results obtained from several sources (UMAE et not UMAE qualified) 3. matrix and vector used only for testing purposes (not derived from a physical process) are derived to evaluate the uncertainty connected with a generic time trend calculated by the code. The present paper describes an application of matrix and vector at item 3. (author)

  14. RELAP5 analysis of integral test of the loss-of-RHR event during the mid-loop operation

    International Nuclear Information System (INIS)

    We are developing a statistical safety evaluation method using the RELAP5/MOD3.2 code for the loss-of-RHR (Residual Heat Removal) event during the mid-loop operation. To confirm the code prediction performance for gravity injection which is a one of the mitigation measures for this event, the Bethsy 6.9a test was analyzed using RELAP5/MOD3.2. In the analytical results, water mass flow rate into the pressurizer in the early period of the transient event was overpredicted. But water mass flow rate into the pressurizer was able to be decreased by artificially giving the circulation flow between core and the core bypass region. (author)

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

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. Implementation of reactor safety analysis code RELAP5/MOD3 and its vectorization on supercomputer FACOM VP2600

    International Nuclear Information System (INIS)

    RELAP5/MOD3 is an advanced reactor safety analysis code developed at Idaho National Engineering Laboratory (INEL) under the sponsorship of USNRC. The code simulates thermohydraulic phenomena involved in loss of coolant accidents in pressurized water reactors. The code has been introduced into JAERI as a part of the technical exchange between the JAERI and USNRC under the ROSA-IV Program. First, the conversion to FACOM (= FUJITSU) M-780 version was carried out based on the IBM version extracted from the original INEL RELAP5/MOD3 source code. Next, the FACOM version has been vectorized for efficient use of new supercomputer FACOM VP2600 at JAERI. The computing speed of vectorized version is about three times faster than the scalar. The present vectorization ratio is 78%. In this report, both the implementation and vectorization methods on the FACOM computers are described. (author)

  18. Analysis of an extreme loss of coolant in the IPR-R1 TRIGA reactor using a RELAP5 model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia Amelia de Lima; Costa, Antonella Lombardi; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Soares, Humberto Vitor, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: hvs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Mesquita, Amir Zacharias, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-07-01

    The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPRR1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits. (author)

  19. 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.

  20. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    International Nuclear Information System (INIS)

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  1. An assessment of the annular flow transition criteria and interphase friction models in RELAP5/MOD2

    International Nuclear Information System (INIS)

    An assessment of the annular flow transition criteria and interphase friction models for two-phase flow in tubes used in RELAP5/MOD2 code is described. The assessment examines the theoretical bases for the criteria and models and considers the results of comparisons with experimental data. Several deficiencies in the transition criteria are identified and appropriate improvements proposed. The interphase friction models are found to be adequate for PWR analyses. (author)

  2. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    Science.gov (United States)

    Sabundjian, Gaianê; Andrade, Delvonei A.; Belchior, Antonio, Jr.; da Silva Rocha, Marcelo; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; de Souza Lima, Ana Cecília

    2013-05-01

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  3. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

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

    International Nuclear Information System (INIS)

    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)

  5. RELAP5/MOD3.2 post-test analysis and CIAU uncertainty evaluation of LOFT experiment L2-5

    International Nuclear Information System (INIS)

    Full text of publication follows: The paper deals with the activity performed at University of Pisa in the framework of the participation to the Phase II and III of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity.The Phase II of BEMUSE Programme is connected with the reanalysis of the Experiment L2 -5 performed in the LOFT (Loss Of Fluid Test) facility in June 1982. The LOFT facility, installed at the Idaho National Engineering Laboratory (INEL), is a 50 MWth Pressurized Water Reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT is typical of large (∼1000 MWe) commercial PWR. For the performance of Experiment L2-5, the LOFT facility was configured to simulate a double-ended 200 % cold leg break in a four-loop PWR operating at nominal conditions. Assumption of loss of offsite power and atypical primary coolant pump coast down were incorporated into the simulation to create core flow stagnation. The light water reactor transient analysis code Relap5/Mod3.2 has been used to simulate this experiment and the standard procedure adopted at

  6. Interpretation of TRIGA reactor experimental data with SIMMER-III code for RELAP5 model evaluation and transient analysis

    International Nuclear Information System (INIS)

    The experimental data obtained in a test campaign carried out in the TRIGA reactor of ENEA/Casaccia are being used to validate a coupled RELAP5/PARCS numerical model for the thermal-hydraulic and dynamic simulation of the reactor. The tests conducted at different power levels provided the temperature measurements at the core inlet and outlet that allow the evaluation of DT through the core at different radial positions. In order to interpret the experimental data for the evaluation of total water mass flow rate through the core in natural circulation, several calculations have been performed with the SIMMER-III code (CFD two-dimensional code) at different core power levels trying to reproduce the experimental measurements. The results of SIMMER-III code were used to fit and validate the simplified 1-D model of the RELAP5 code used for thermal-hydraulic transient analysis of TRIGA reactor. Finally, this paper presents the interpretation of some reactivity transients using this improved T/H model with the RELAP5 point-kinetics neutronic model. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. RELAP5 calculations show that higher gas flow rates are required to initiate the flooding compared with the experimental data if the L/D is as low as that of the hot legs of typical PWRs. Based on the present data bank, the new CCFL correlation is derived, which shows the L/D effect. The present correlation agrees well with the database within the prediction error, 8.7% and it is implemented into the RELAP5 and validated against the data bank. The predictions of the flooding limit by the modified version lie well on the applied CCFL curve if the L/D is lower than 22, which is the case of the hot legs of typical PWRs

  9. 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.

  10. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

    Directory of Open Access Journals (Sweden)

    P. Balestra

    2016-01-01

    Full Text Available The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential, specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead, were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

  11. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  12. WWER-1000 thermal /hydraulic model for determining boundary conditions for fracture toughness assessment with use of RELAP5/mod3.2 computer code

    International Nuclear Information System (INIS)

    The paper describes changes in the WWER-1000 reactor model for RELAP5/mod3.2 computer code to permit more detailed modeling of the downcomer by its azimuthal division into 20 vertical channels with cross-flow junctions

  13. Applicability research of RELAP5/MOD3.3 for small break loss of coolant accident of NPP with passive safety system

    International Nuclear Information System (INIS)

    The passive core cooling system is used in AP1000 to mitigate the small break loss of coolant accident (SBLOCA). The RELAP5/MOD3.3 code is generally applicable to the traditional NPP SBLOCA research, but for the passive NPP SBLOCA, its applicability will need further study and evaluation. Based on the analysis of the important phenomenon of the SBLOCA of the passive NPP, the RELAP5/MOD3.3 code was assessed and modified. In order to verify the applicability of the modified RELAP5/MOD3.3 code, the DBA-02 and NRC-05 cases of APEX-1000 which was the test facility for verifying AP1000 small break loss of coolant accident, were simulated. It shows good agreement between the results of the modified RELAP5/MOD3.3 code and experiment data. (authors)

  14. SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

    International Nuclear Information System (INIS)

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head

  15. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

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

    International Nuclear Information System (INIS)

    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

  17. A post test analysis of the OECD LOFT experiment LP-FP-2 using the computer programs SCDAP/RELAP5, TRAP-MELT2.2 and PULSE

    International Nuclear Information System (INIS)

    A post-test analysis has been performed to access the current state of early severe accident models within the SCDAP/RELAP5, TRAP-MELT2.2 and PULSE computer programs. The experimental data base against which the severe accident models were assessed was obtained during the OECD LOFT LP-FP-2 experiment. The experiment was performed under the auspices of the OECD-LOFT consortium. This consortium consisted of representatives from 10 nations. The LOFT LP-FP-2 experiment was performed in the Loss-Of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory (INEL) in 1985. This experiment had the objective of obtaining fission product release, transport, and deposition data during the early phases of a risk dominant reactor transient to establish a benchmark data base for (i) assessing the understanding of the physical phenomena controlling reactor system fission product behavior, and (ii) assessing the capability of computer models to predict reactor system fission product release and transport. The analyses described in this report assess the thermal-hydraulic, core relocation, and fission product transport models currently available to reactor safety analyses. SCDAP/RELAP5 was assessed for its ability to model the thermal-hydraulic and core relocation behavior. The TRAP-MELT and PULSE codes were assessed for their ability to model fission product transport and deposition once boundary conditions had been established. The computer programs are obtained from the United States within the Swiss-United States Nuclear Regulatory Commission (USNRC) agreement on the severe accident research program. The SCDAP/RELAP5 and PULSE codes have been developed at the Idaho National Engineering Laboratory and TRAP-MELT2.2 code at the Battelle Columbus Laboratory. (author) 18 figs., 7 tabs., 9 refs

  18. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 3. Application of RELAP5-3D and RELAP5/ MOD3.22 to Phase I of OECD PWR MSLB Benchmark

    International Nuclear Information System (INIS)

    same, except the initial steam generator mass inventory and the OTSG outlet temperature. The RELAP5-3D predicts a higher value of the mass in the steam generator, i.e., 28 094 kg, compared to the RELAP5/MOD3.22 calculation of 26 220 kg (Fig. 1). The reference value of the specifications is 28 395 kg. An important root cause of this difference is the different interfacial drag model used in the codes. In addition, a large initial fluid mass in the broken steam generator enhances the potential primary cooling system capability. The consequences of the calculation of the transient are mostly connected with the return-to-power (RTP) phenomenon (Fig. 2), where the total power versus time is represented. Although the first power peak has similar features for the two code runs, the second power peak predicted by RELAP5-3D is higher and shows a well-defined shape compared to the RELAP5/MOD3.22 calculation that is characterized by two peaks. Moreover, the core power after the second peak converges to a similar value at a similar time in both calculations. The other parameters that characterize the transient have similar times and tend mostly to coincide. The main conclusions of the analysis of the MSLB can be summarized as follows: 1. The adopted nuclear power plant (TMI-1) is safe enough in the considered scenario. 2. Both codes have predicted scram and RTP phenomena. The results obtained are qualitatively similar; however, in quantitative terms, noticeable differences have been found. In particular, 1. The initial steam generator mass and consequently the RTP peak are influenced by different thermal-hydraulic models governing the interfacial drag. 2. Concerning the different shape of the second power peak, one must notice that the beta version of RELAP5/MOD3.22 was adopted. Further calculations with the gamma version of RELAP5/MOD3.22, have shown that the second power peak is similar to that obtained with the RELAP5-3D code. 3. Because of the sensitive nature of this

  19. Modeling of control rod ejection transient for WWER-1000-model 446 using RELAP5m3.3/PARCSv2.6 coupled codes

    International Nuclear Information System (INIS)

    Highlights: • Capability to perform 3D neutronics/thermal–hydraulic analysis for WWER-1000 m446. • Good agreement was observed between the coupled codes results and FSAR data. • Capability to perform multi-dimensional analysis of complex transients such as a CREA. • WWER-1000 m446 shows a safe response during these transients performance. - Abstract: By using the Best Estimate (BE) method instead of conservative assumptions for the evaluation of reactor safety, significant economic considerations with optimal fuel burn-up could be obtained in addition to reactor safety. In this method, due to the detailed simulation and feedback considerations, special attention has been paid to the coupling of neutronic and thermo-hydraulic codes to achieve more reliable results. In this study, the Control Rod Ejection (CRE) transient has been simulated for Bushehr Nuclear Power Plant (BNPP) as a WWER-1000 power plant model 446 according to Final Safety Analysis Report (FSAR). CRE is a transient of Reactivity Initiated Accidents (RIA) category. In this study, the reactor thermo-hydraulic system has been simulated by RELAP5/mod3.3, while the neutron kinetic system of the reactor core has been simulated by the PARCSv2.6 code. These codes have been coupled utilizing Parallel Virtual Machine (PVM) interface software to consider the effects of thermal hydraulic and neutronic feedbacks. Thus, the power calculated by the PARCS code is used by the RELAP5 code and the obtained thermal hydraulic parameters are inserted to the PARCS code for macroscopic cross-section calculations. A computer program written by C++ has been used for the cycle execution of the WIMS code to produce the macroscopic cross-section library with the format required by the PARCS code. After the three-dimensional (3D) thermo-neutronic modeling of the reactor core, the Hot Zero Power (HZP) and Hot Full Power (HFP) versions of CRE transients, which have been considered in the plant’s FSAR, have been

  20. Analysis fom the OECD/NEA PWR Main Steam Line Break (MSLB) Benchmark exercise 3 with the Coupled Code System RELAP5/PANBOX

    International Nuclear Information System (INIS)

    The main purpose of the computational OECD/NEA PWR MSLB-Benchmark is the evaluation of the prediction capability of advanced code systems by means of a code-to-code comparison. The postulated MSLB-transient is characterized by a strong non-symmetrical core thermal behaviour due to the feedback between neutron kinetics and plant thermal hydraulics. The analysis of such transients with pronounced spatial power distortion represents a considerable challenge for advanced code systems. It is initiated by a break of one main steam line when the reactor TMI-1 is operated at nominal power. High heat removal through the break leads to a strong cooldown rate of the broken loop compared to the intact one. Under such conditions a power increase and a re-criticality of the core despite scram can not be excluded due to the negative reactivity coefficients. The MSLB-Benchmark enfolds three exercises as follows: Exercise 1: integral plant simulation with best-estimate codes using the point kinetics, Exercise 2: multidimensional simulation of the core for given initial and boundary conditions, and Exercise 3: integral plant simulation with coupled, best-estimate codes using 3D-neutron kinetics models. Forschungszentrum Karlsruhe (FZK) and Framatome advanced nuclear power (ANP) Erlangen participated on the MSLB-Benchmark with the code system RELAP5/PANBOX for the Exercise 3: Based on the plant and core models elaborated for Exercise 1 and 2, an integral TMI-1 plant model was elaborated for Exercise 3. Special emphasis was put on the development of a multidimensional core model for the space-time kinetics. Two scenarios, the best-estimate (BE) and the return-to-power (RP) scenario, were investigated. Additional investigations aimed to investigate the influence of the coolant mixing on re-criticality and power increase. Results of these investigations are presented and discussed in this report. It has been demonstrated that RELAP5/PANBOX is capable to simulate complex transient in a

  1. RELAP5/PARCSV2.7 qualification for BWR stability simulations. Application to peach bottom NPP

    International Nuclear Information System (INIS)

    To characterize the unstable behavior of the Peach Bottom Unit 2 BWR, a number of perturbation analyses were performed: arrangements with Philadelphia Electric Company (PECo) were made for conducting different series of Low Flow-Stability Tests at Peach Bottom 2, during the first quarter of 1977. The Low Flow Stability Tests intended to measure the reactor core stability margins at the limiting conditions used in design and safety analysis, providing a one-to-one comparison to design calculations. Stability tests were conducted along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. In this work, three dimensional time domain BWR stability analysis were performed on a new analysis point (PTUPV), which is inside the exclusion region with a core mass flow of 4660.1 kg/s (34% of the core rated mass flow) and total reactor power of 1997.8 MW (60.7 of the core rated reactor power), using the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by the control rod movement as it is usual performed in Nuclear Power Plants. For the core, 48 thermalhydraulic channels have been modeled to represent the active part of the core and one channel for all by-passes. The thermalhydraulic-to-neutronic mapping has been made based on the fundamental and first and second harmonics shapes of the reactor power, calculated with the modal code LAMBDA. For the rest of the plant a coarse nodalization has been adopted for limiting the needed computer resources. For the neutronic code, a nodalization with a 3D core mesh composed with 764 axial nodes has been modeled. A large set of cross section data including 435 compositions has been adopted in neutronic input deck. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. The calculated results show that point PTUPV is an unstable point and the obtained

  2. Influence of the thermalhydraulic to neutronic channel mapping in a 3D rea analysis with RELAP5/PARCS v2.7 at Trillo NPP

    International Nuclear Information System (INIS)

    The progress in analytical methods has evolved the classical thermalhydraulic codes such as RETRAN, TRAC and RELAP5 towards modern codes with full capability of performing 3D kinetics analyses in a dynamic way, for simulating the behaviour of specific cores in a realistic manner and so to predict the localized power excursions as occurs in the RIA. These codes must be feed with the kinetics information of physics codes like CASMO4-SIMULATE3. SIMTAB methodology provides an easy tool for properly extracting and formatting the cross-sections and neutronic kinetic parameters from SIMULATE to the coupled neutronic-thermalhydraulic codes, making feasible reactivity-based studies in BWR and PWR cores. SIMTAB allows to accurately transfer the initial kinetic status of the core from the physics code to the thermalhydraulic code and provides the adequate kinetics response during the full transient. We have analyzed the behavior of the Trillo NPP core in a REA with the coupled neutronic-thermalhydraulic code RELAP5/PARCS v2.7 using the cross-section set and other kinetic parameters obtained with the application of the SIMTAB methodology, developed in UPV. We have study this transient in different operating conditions and at the beginning and at the end of cycle. The present work consists of the study of the influence of different definitions of the thermalhydraulic model in a REA analysis at Trillo NPP. A series of calculations with different number of thermalhydraulic channels to represent the core has been made. These channels have been coupled to the neutronic model, developed in a one-to-one basis, that is, each fuel assembly is represented by a radial node in PARCS V2.7 code. The mapping between the thermalhydraulic and the neutronic model has been performed in different ways to study its influence in the 3D results. The results have shown that the power peak reached in this transient depends strongly on the core thermalhydraulic model. Furthermore, the axial power

  3. Simulation of a simple RCCS experiment with RELAP5-3D system code and computational fluid dynamics computer program

    International Nuclear Information System (INIS)

    A small scale experimental facility was designed to study the thermal hydraulic phenomena in the Reactor Cavity Cooling System (RCCS). The facility was scaled down from the full scale RCCS system by applying scaling laws. A set of RELAP5-3D simulations were performed to confirm the scaling calculations, and to refine and optimize the facility's configuration, instrumentation selection, and layout. Computational Fluid Dynamics (CFD) calculations using StarCCM+ were performed in order to study the flow patterns and two-phase water behavior in selected locations of the facility where expected complex flow structure occurs. (author)

  4. Analysis of integral circulation and decay heat removal experiments in the lead-bismuth CIRCE facility with RELAP5 code

    International Nuclear Information System (INIS)

    In this paper, the results of the post-test analysis of some integral circulation experiments conducted on the lead-bismuth CIRCE facility are presented in comparison with the experimental data. These experiments include the simulation of unprotected loss of flow and unprotected loss of heat sink transients in a pool-type heavy liquid metal reactor. Furthermore, the results of the pre-test analysis of a protected loss of heat sink and flow transient with decay heat removal by a heat exchanger immersed in the pool and operating in natural circulation is presented. All transient analyses have been performed with the RELAP5 thermal-hydraulic code. (author)

  5. RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

    International Nuclear Information System (INIS)

    Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available

  6. Experiences with the coupled code system S3R/RELAP5-3D in training simulators

    International Nuclear Information System (INIS)

    The paper describes the implementation of S3R (core neutronics) and RELAP5-3D (RCS thermal-hydraulics) in a PWR training simulator for the Grohnde NPP located at the Kraftwerks-Simulator-Gesellschaft (KSG) in Essen, Germany. The models are briefly described as well as the coupling between these two codes and the interface with the rest of the simulator. The paper also describes the procedure that will be used to update the simulator core data after future core reloads. Results from the integrated simulator are presented. (orig.)

  7. Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests

    International Nuclear Information System (INIS)

    The CCFL model used in RELAP5/MOD3 version 5m5 has been assessed against simple vertical tubes and bundle tests performed at a facility of Korea Atomic Energy Research Institute. The effect of changes in tube diameter and nodalization of tube section were investigated. The roles of interfacial drags on the flooding characteristics are discussed. Differences between the calculation and the experiment are also discussed. A comparison between model assessment results and the test data showed that the calculated value lay well on the experimental flooding curve specified by user, but the pressure jump before onset of flooding was not calculated

  8. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  9. Developmental assessment of RELAP5/MOD3.1 with separate-effect and integral test experiments: model changes and options

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.T. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-04-01

    A summary of modifications and options introduced in RELAP5/MOD3.1 (R5M3.1) is presented and it is shown that the predicting capabilities of the modified version of the code are greatly improved, while the general philosophy we followed in arriving at these modifications is also outlined. These changes which are the same ones we implemented in the past in the version 7j of the code, include 2 different heat transfer packages (one of them activated during reflooding), modification of the low mass-flux Groeneveld CHF look-up table and of the dispersed flow interfacial area (and shear) as well as of the criterion for transition into and out from this regime, almost complete elimination of the under-relaxation schemes of the interfacial closure coefficients etc. The modified R5M3.1 code is assessed against a number of separate-effect and integral test experiments and in contrast to the frozen version, is shown to result in physically sound predictions which are much closer to the measurements, while almost all the predicted variables are free of unphysical spurious oscillations. The modifications introduced solve a number of problems associated with the frozen version of the code and result in a version which can be confidently used both for SB-LOCA and LB-LOCA analyses. (author) 7 figs., 20 refs.

  10. Analysis of small-break loss-of-coolant accident test 9.1b at BETHSY facility with TRACE and RELAP5

    International Nuclear Information System (INIS)

    Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. The TRAC/RELAP Advanced Computational Engine (TRACE) is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission. The purpose of the present study was to assess the accuracy of the TRACE calculation of BETHSY 9.1b test comparing to RELAP5 calculation. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck and then some adaptations were needed too. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 3 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. In general, the TRACE code calculation is in good agreement with the BETHSY 9.1b test. The calculation results are as good as or better than the RELAP5 calculated results until low pressure injection period. It should be mentioned that the RELAP5 developed animation model was of great help in investigating the calculated physical phenomena and preparing the preliminary TRACE analysis. (authors)

  11. Total Transfer Capability Assessment Incorporating Corrective Controls for Transient Stability using TSCOPF

    Science.gov (United States)

    Hakim, Lukmanul; Kubokawa, Junji; Yorino, Naoto; Zoka, Yoshifumi; Sasaki, Yutaka

    Advancements have been made towards inclusion of both static and dynamic security into transfer capability calculation. However, to the authors' knowledge, work on considering corrective controls into the calculation has not been reported yet. Therefore, we propose a Total Transfer Capability (TTC) assessment considering transient stability corrective controls. The method is based on the Newton interior point method for nonlinear programming and transfer capability is approached as a maximization of power transfer with both static and transient stability constraints are incorporated into our Transient Stability Constrained Optimal Power Flow (TSCOPF) formulation. An interconnected power system is simulated to be subjected to a severe unbalanced 3-phase 4-line to ground fault and following the fault, generator and load are shed in a pre-defined sequence to mimic actual corrective controls. In a deregulated electricity market, both generator companies and large load customers are encouraged to actively participate in maintaining power system stability as corrective controls upon agreement of compensation for being shed following a disturbance. Implementation of this proposal on the actual power system operation should be carried out through combining it with the existing transient stabilization controller system. Utilization of these corrective controls results in increasing TTC as suggested in our numerical simulation. As Lagrange multipliers can also describe sensitivity of both inequality and equality constraints to the objective function, then selection of which generator or load to be shed can be carried out on the basis of values of Lagrange multipliers of its respective generator's rotor angle stability and active power balance equation. Hence, the proposal in this paper can be utilized by system operator to assess the maximum TTC for specific loads and network conditions.

  12. Experimental and analytical study to improve noncondensible gas model of RELAP5 code for small and medium integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Un Chul; Park, Gun Chul; Suh, Kyun Ryul; Lee, Seung Wook; Suh, Jung Kwan [Seoul National University, Seoul (Korea); Kim, Mu Hwan [Pohang Science and Technology University, Pohang (Korea); Kim, Sin [Cheju National University, Cheju (Korea)

    1999-03-01

    In this study, the effect of noncondensibles dissolved and effervescent in reactor coolant was surveyed and the transport and effervescence model was developed. And the modification of RELAP5 code and the verification test were performed. Details are followings; 1. Examination of the code and the design of test facilities - Examination of the structure and analysis model of the code - Verification of the noncondensible gas model and boron transport model in the code - Design of noncondensible test facilities 2. Development of models and the manufacturing the test facilities - Introducing transport model of noncondensible solubles - Development of noncondensible effervescence model - Manufacturing the test facilities and verification of the facilities 3. Verification of the modified code - Establishment of natural circulation heat transfer model in the presence of noncondensible soluble - Modification of RELAP5 heat transfer model - Verification of the modified code. The result deduced from the experiment will be utilized the basic data for the future experiment of plant operation conditions, and the models for noncondensible soluble can be applied to the developing of new thermal-hydraulic computer code. (author). 20 refs., 65 figs., 12 tabs.

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

    Directory of Open Access Journals (Sweden)

    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.

  14. Conversion of the thermal hydraulics components of Almaraz NPP model from RELAP5 into TRAC-M

    International Nuclear Information System (INIS)

    In the scope of a joint project between the Spanish Nuclear Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) on the USNRC state-of-the-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper refers to the translation of the following thermal hydraulic models: pressurizer, hot and cold legs (including the pumps and the injection systems), and steam generators. The comparison of the results obtained with both codes shows a good agreement. However, some difficulties have been found in the translation of some code components, like pipes, heat structures, pumps, branchs, valves and boundary conditions. In this paper, these translation problems and their solutions are described.(author)

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

    International Nuclear Information System (INIS)

    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

  16. Discussion on RELAP5 and RETRAN3D Modeling for Passive Condensate Cooling Tank of Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    Domestic nuclear industry has started the development of APR+ as a Korean specific reactor for the export strategy. In the development of APR+ a passive auxiliary feedwater system (PAFS) has been considered as a noticeable candidate of improved design. The outline of PAFS and passive condensate cooling tank (PCCT) containing horizontal heat exchanger is shown in Fig. 1. For the successful design of PAFS, performance analyses or safety analyses are prerequisite using best estimate thermal hydraulic codes such as RELAP5 or RETRAN3D. Because of the inherent features of RELAP5 or RETRAN3D, pool model and condensation in horizontal tube have not been well-setup nor widely studied. This paper discusses about the PCCT phenomena including steam condensation in horizontal tube and pool heat transfer, and RELAP5 and RETRAN3D modeling

  17. Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes

    International Nuclear Information System (INIS)

    RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)

  18. Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y.; Rohde, U. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Baranaev, Y.; Glebov, A. [State Scientific Center of the Russian Federation, Obninsk, Kaluga Region (Russian Federation). Inst. for Physics and Power Engineering

    2012-08-15

    RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)

  19. Qualification of RELAP5/MOD3 for safety relief valve hydrodynamic load analysis: A comparison against EPRI/CE SRV test 1017

    International Nuclear Information System (INIS)

    This paper documents the acceptability of the default two-velocity momentum equation option in the RELAP5/MOD3 computer program for the estimation of hydrodynamic loads associated with steam safety relief valve discharge. A RELAP5 analysis of the EPRI/Combustion Engineering Safety Valve Test Loop Facility was performed. Time-dependent hydrodynamic forcing functions for the four pipe segments of the Combustion Engineering Test Facility were developed. These forcing functions were subsequently used in an elastic piping analysis model to estimate the resultant structural responses. The calculated loads were then compared to the values from the original 1981 Test data (Dresser safety valve Test 1017 with cold water loop seal). The results verify that RELAP5/MOD3 and the REFORC post-processor can be used with confidence to calculate hydrodynamic forces for use in pipe stress and support analysis

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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)

    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)

  3. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 2. Sensitivity Studies for MSLB Exercises 2 and 3 with RELAP5/PANBOX

    International Nuclear Information System (INIS)

    As a contribution to the verification and validation of the RELAP5/PANBOX coupled code system (R/P/C), we took part in the Main-Steam-Line-Break (MSLB) Benchmark issued by OECD/NEA. Sensitivity studies with respect to external/ internal integration and coarse/fine channel representation have already been presented for exercise 2. The purpose of this paper is to extend the sensitivity studies to exercise 3 also and to present local results for safety-related parameters. R/P/C is a nuclear plant safety analysis code system that consists of the PANBOX core simulator coupled to the RELAP5 best-estimate plant simulator. The coupling is performed via the EUMOD RELAP5 interface package. R/P/C has the capabilities of RELAP5 with added ability for calculation of three-dimensional (3-D) neutronics and thermal margins with COBRA, the core thermal-hydraulic module of PANBOX. The neutronics nodalization is radially based on one node per fuel assembly (FA). Axially, 28 layers are modeled, where the specified mesh sizes are used with the exception of the 2 layers of 29.76 cm, which are subdivided into 4 layers. All calculations use the semi-analytical Nodal Expansion Method. The time discretization is based on the implicit Euler method combined with the exponential transformation technique. In the external integration of R/P/C, the core thermal-hydraulics solution is calculated by COBRA using core inlet boundary conditions from RELAP5. The channel geometry is based on one channel/FA, with axially 24 layers. In the internal integration, the core thermal-hydraulics solution is calculated by RELAP5. The channel geometry is based on 19 coarse channels with axially 11 core layers. R/P/C allows hot subchannel analysis by application of an on-line refinement of channels (HOSCAM). Fuel assembly powers, hot pin powers, and powers of a surrounding subchannel region are passed to COBRA for selected FAs in the external integration option. COBRA performs subchannel analysis by using a

  4. Simulation of a channel blockage transient in the Angra 2 Nuclear Reactor using a RELAP5-3D model

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Mantecon, Javier; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia; Reis, Patricia A.L.; Scari, Maria E., E-mail: mantecon1987@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: patricialire@yahoo.com.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The Angra 2 Nuclear Power Plant (NPP) is a Pressurized Water Reactor (PWR) type with electrical output of about 1350 MW. The RELAP5-3D code was used to develop a detailed thermal hydraulic model of such reactor using reference data from the Angra 2 Final Safety Analysis Report (FSAR). In this work, a blockage transient has been investigated at full power operation. The transient herein considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  5. Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

    International Nuclear Information System (INIS)

    This research aims to develop reliable, advanced system thermal-hydraulic computer code and to quantify the uncertainties of code to introduce the best estimate methodology of ECCS for LBLOCA. Although the one of best estimate code, RELAP5/MOD3.1 was introduced from USNRC, several deficiencies in its reflood model and some improvements have been made. The improvements consist of modification of reflood wall heat transfer package and adjusting the drop size in dispersed flow regime. The tome smoothing of wall vaporization and level tracking model are also added to eliminate the pressure spike and level oscillation. For the verification of improved model and quantification of associated uncertainty, the FLECHT-SEASET data were used and upper limit of uncertainty at 95% confidence level is evaluated. (Author) 30 refs., 49 figs., 2 tabs

  6. Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Chung, Bob Dong; Lee, Young Jin; Park, Chan Eok; Lee, Guy Hyung; Choi, Chul Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This research aims to develop reliable, advanced system thermal-hydraulic computer code and to quantify the uncertainties of code to introduce the best estimate methodology of ECCS for LBLOCA. Although the one of best estimate code, RELAP5/MOD3.1 was introduced from USNRC, several deficiencies in its reflood model and some improvements have been made. The improvements consist of modification of reflood wall heat transfer package and adjusting the drop size in dispersed flow regime. The tome smoothing of wall vaporization and level tracking model are also added to eliminate the pressure spike and level oscillation. For the verification of improved model and quantification of associated uncertainty, the FLECHT-SEASET data were used and upper limit of uncertainty at 95% confidence level is evaluated. (Author) 30 refs., 49 figs., 2 tabs.

  7. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    International Nuclear Information System (INIS)

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions

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

    International Nuclear Information System (INIS)

    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

  9. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  10. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3DC/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

  12. Assessment study of RELAP5/MOD2, CYCLE 36. 04 based on spray start-up test for DOEL-4

    Energy Technology Data Exchange (ETDEWEB)

    Moeyaert, P.; Stubbe, E.

    1989-07-01

    This report presents an assessment study for the code RELAP-5 MOD-2 based on a pressurizer spray start-up test of the Doel-4 power plant. Doel-4 is a three loop WESTINGHOUSE PWR plant ordered by the EBES utility with a nominal power rating of 1000 MWe and equipped with preheater type E steam generators. A large series of commissioning tests are normally performed on new plants, of which the so called pressurizer spray and heater test (SU-PR-01) was performed on February 2nd 1985. TRACTEBEL, being the Architect-Engineer for this plant was closely involved with all start-up tests and was responsible for the final approval of the tests.

  13. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  14. Assessment of TRAC-PF1 and RELAP5/MOD1 codes with GE large-vessel blowdown test

    International Nuclear Information System (INIS)

    The GE large vessel blowdown Test No. 5801-15 was simulated with the TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes. The test facility consisted of a pressure vessel, 49-in in diameter by 14-ft long, a 2.5-in diameter converging-diverging nozzle and a blowdown line connected to the center of the upper part of the vessel (elevation from the bottom of the vessel 10.5 ft). The vessel was filled with saturated water up to 5.5 ft at 1060 psia. The test was initiated by rupturing a disc attached at the end of the nozzle. The purpose of this experiment was to study blowdown phenomena such as critical blowdown flow and the level swell during blowdown from a partially water filled vessel. Understanding of these phenomena is essential for the analysis of Loss-of-Coolant (LOCA) and steam generator steam line break accidents

  15. 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.

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

    International Nuclear Information System (INIS)

    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

  17. Long term station black out analyses for TAPS - 3 and 4 with RELAP5/MOD.3.2

    International Nuclear Information System (INIS)

    A postulated situation involving failure of both Class-IV and Class-III power supplies is referred to as Station Black Out (SBO) condition is analyzed for Tarapur Atomic Power Station-3 and 4 (TAPS- 3 and 4). During Station Black out, PHT system PHT circulation will be initially provided by the coastdown of PHT main circulating pump flywheels and later by thermosyphoning. On the secondary side, fast depressurization has been achieved through crash cool down and afterheat removal by Fire Fighting Water Pumps through steam generators. Thermal hydraulics code RELAP5/MOD3.2 has been used for this long term station black out analyses (72 hours) and are discussed in the paper. (author)

  18. Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Nathan V. Hoffer; Piyush Sabharwall; Nolan A. Anderson

    2011-01-01

    Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.

  19. Transient simulations in WWER-1000-comparison between DYN3D-ATHLET and DYN3D-RELAP5

    International Nuclear Information System (INIS)

    Simulations of a real transient of an operating WWER-1000 power plant have been performed using DYN3D-ATHLET (Gru95) and DYN3D-RELAP5 (Koy01) code systems in the frame of activities aimed at a validation of the neutronic / thermal-hydraulic coupled codes. The transient initiated by a main coolant pump switching off, when three of the four main coolant pumps of the plant are in operation (scenario of the VALCO project) is chosen for the simulation. The same models of the plant (except the core nodalization) but two different libraries of macroscopic cross-sections have been used in compared calculations. Additionally, the compared code systems are based on the different / external and internal / coupling techniques. This paper contains a brief description of the coupled codes and the plant model as well as a comparison between the results from simulations (Authors)

  20. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    International Nuclear Information System (INIS)

    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. 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. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours 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. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  1. RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization

    International Nuclear Information System (INIS)

    RELAP5 code post-test analyses were performed on two ROSA/large scale test facility (LSTF) validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using steam generator (SG) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs a little after a safety injection signal. Auxiliary feedwater injection was done into the secondary-side of both SGs thereafter. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by boil-off. Core collapsed liquid level recovered after the initiation of accumulator (ACC) coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection (LPI) system. In the 4-in. break test, on the other hand, no core uncovery and heatup happened due to the coolant injection from the ACC and LPI systems. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code predicted well the overall trend of the major thermal-hydraulic response observed in the two LSTF tests. The code, however, overpredicted the peak cladding temperature (PCT) because of underprediction of the core collapsed liquid level due to inadequate prediction of the ACC flow rate in the 8-in. break case. Sensitivity analyses with the RELAP5 code indicated that a time delay for the SG depressurization start and break discharge coefficient for two-phase discharge flow affect the PCT significantly in the 8-in. break case. (author)

  2. Peach bottom instability analysis with a RELAP5/PARCSv2.7 detailed thermal-hydraulic–neutronic model

    International Nuclear Information System (INIS)

    Highlights: ► A RELAP5-MOD3.3/PARCSv2.7 model developed to characterize oscillations in BWR. ► The TH to neutronic mapping is based on the Lambda modes obtained with VALKIN code. ► The results show that point PTUPV is an unstable point with a bottom-peaked profile. ► The SSA and the Power Modal Decomposition have been applied to the LPRMs data. ► An in-phase coupled with an out-of-phase oscillation appears. - Abstract: In this work, BWR stability analysis has been performed on an operating point (PTUPV) of Peach Bottom NPP which is inside the exclusion region in the operating power-flow map. The simulation has been made with the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by a control rod movement as it is usually performed in Nuclear Power Plants. The thermal-hydraulic model is a detailed model that includes all the reactor vessel components: jet pumps, recirculation pumps, downcomer, reactor core and also the separator and the dryer. The reactor core has been modeled with 72 thermal-hydraulic channels, 71 represent the active core and 1 represents the core bypass. The reactor core thermal-hydraulic to neutronic representation (mapping) has been divided in four quadrants according to the first and second power harmonics (Lambda modes) obtained previously with the VALKIN code. This mapping was chosen in order not to condition the oscillation pattern. The transient starts with the control rod movement. The calculated results show that point PTUPV is an unstable point and the obtained relative axial power distribution shows a bottom-peaked profile, which is characteristic of unstable cores.

  3. Analysis of experiments for in-tube steam condensation with noncondensable gas at low pressure using the RELAP5/MOD3.2 code modified with a non-interactive condensation model

    International Nuclear Information System (INIS)

    The standard RELAP5/MOD3.2 code is modified using non-interactive modeling, which is a mechanistic model developed for easy engineering application to simulate steam condensation in the presence of noncondensable gases in a tube. To predict the liquid-side heat transfer coefficients in the modified RELAP5/MOD3.2 code, Nusselt's correlation is used for condensation in a vertical tube and Kim's correlation correlated with the Froude number is used for condensation in a horizontal tube. In the modified code the wall friction in a vertical tube is calculated using the two-phase friction factor correlation proposed by Collier and the interfacial friction factor is calculated using the empirical power-law relationship of Choi. Both the standard and the modified RELAP5/MOD3.2 codes are used to simulate two kinds of vertical in-tube experiments and a horizontally stratified in-tube experiment involving the condensation phenomenon in the presence of noncondensable gas. Two vertical in-tube experiments, PCCS (Passive Containment Cooling System) condensation and reflux condensation experiments, provide data on the steady-state behaviors with typical flow, pressure and air mass fraction conditions likely to be seen in a condensing tube of PCCS and in a U-tube of steam generator in-mid-loop operation. The horizontally stratified in-tube experiment represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard codes for steam condensation in the presence of noncondensable gas are assessed using those three KAIST condensation experiments. The modified code gives better prediction over the data of the three condensation experiments than the standard code. Simulation results of PCCS and reflux condensation experiments show that the local heat transfer coefficients are well predicted with the modified code but they are under-predicted by the default model and over-predicted by the

  4. Final results of the sixth three-dimensional AER dynamic Benchmark problem calculation. Solution of problem with DYN3D and RELAP5-3D codes

    International Nuclear Information System (INIS)

    The paper gives a brief survey of the 6th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI Rez. This benchmark was defined at the 10th AER Symposium. Its initiating event is a double ended break in the steam line of steam generator No. 1 in a WWER-440/213 plant at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations as well as tuning of initial state before the transient were performed with the code DYN3D. Transient calculations were made with the system code RELAP5-3D. The KASSETA library was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the 6th AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was connected with 37 coolant channels thermal-hydraulic model of the core, 6-sector nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5 -3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters (Authors)

  5. Analysis of the Peach Bottom NPP stability using a core neutronic-thermohydraulic model with RELAP5/PARCS v2.7 connected code

    International Nuclear Information System (INIS)

    The aim of this study is to test the RELAP5/PARCS v2.7 connected code. The results show the pint PTUPV is an unstable point and the axial obtained power distribution shows a pierced profile in the bottom of the core, typical of unstable cores.

  6. Analysis of the OECD/NEA PWR Main Steam Line Break (MSLB) Benchmark Exercise 1 using the RELAP5 code with the point kinetics option

    International Nuclear Information System (INIS)

    The main purpose of the computational OECD/NEA pressurized water reactor main steam line break (PWR MSLB) Benchmark is the evaluation of the prediction capability of advanced code systems by means of a code-to-code comparison. The postulated MSLB-transient is characterized by a strong non-symmetrical core thermal behaviour due to the feedback between neutron kinetics and plant thermal hydraulics. The analysis of such transients with pronounced spatial power distortion represents a considerable challenge for advanced code systems. The transient is initiated by a double-ended break of one main steam line when the reactor TMI-1 is operated at nominal power. The high heat removal through the break leads to a strong cooldown of the primary coolant. Under such conditions a power increase and a re-criticality of the core despite scram can not be excluded due to the negative reactivity coefficients. The MSLB-Benchmark enfolds three exercises as follows: Exercise 1: integral plant simulation with best-estimate codes using point kinetics, Exercise 2: multidimensional simulation of the core for given initial and boundary conditions, and Exercise 3: integral plant simulation with coupled, best-estimate codes using 3D-neutron kinetics models. Das Forschungszentrum Karlsruhe (FZK) and Framatome Advanced Nuclear Power/Erlangen (former Siemens/KWU) participated on the MSLB-Benchmark with the code system RELAP5/MOD3.2 for the Exercise 1: In this report, the integral plant model developed for this Exercise 1 together with the calculated results will be presented and discussed. (orig.)

  7. A comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5/MOD3 and CEFLASH-4AS/REM computer codes

    International Nuclear Information System (INIS)

    The Ulchin 3 and 4 nuclear power plants, which are two-loop 2,825 MW(thermal) pressurized water reactors designed by the Korea Atomic Energy Research Institute, adopted a safety depressurization system (SDS) to mitigate the beyond-design-basis event of a total loss of feedwater (TLOFW). A comparative simulation by the CEFLASH-4AS/REM and RELAP5/MOD3 computer codes for the TLOFW event without operator recovery and the TLOFW event with feed and bleed (F and B) operation is performed for Ulchin 3 and 4. In the analyses, the SDS bleed paths are modeled by orifices located on the top of the pressurizer, where the analytical area of the bleed path is based on the Ulchin 3 and 4 SDS design flow capacity. An additional case, where the SDS piping and valves are modeled explicitly, is considered for the RELAP5 analysis. The predictions by the CEFLASH-4AS/REM of the transient two-phase system behavior show good qualitative and quantitative agreement with those by the RELAP5 simulation. The RELAP5 case with explicit piping results in less repressurization and lower reactor coolant system pressure than that of the case without explicit SDS modeling. However, the two cases of RELAP5 analyses result in essentially the same transient scenarios for TLOFW with F and B operation. The results of the simulation demonstrate the validity of the Ulchin 3 and 4 design approach, which employs CEFLASH-4AS/REM computer code and SDS bleed paths modeled by orifices located on the top of the pressurizer. The results also indicate that the decay heat removal and core inventory makeup function can be successfully accomplished by F and B operation by using the SDS for Ulchin 3 and 4

  8. Safety Analysis on Dual-functional Lithium Lead Test Blanket Module With RELAP5%基于 RELAP5的双功能液态锂铅实验包层模块安全分析

    Institute of Scientific and Technical Information of China (English)

    李伟; 田文喜; 秋穗正; 苏光辉

    2013-01-01

    利用嵌入了液态锂铅(LiPb)的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅(DFLL)实验包层模块(TBM)的安全特性进行评价。对DFLL-TBM 及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁(FW )结构材料表面最高温度低于允许值550℃。事故工况下氦气泄漏引起的ITER真空室(VV)、窗口设备室(port cell)以及托卡马克冷却水系统大厅拱顶(TCWS vault)的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM 的设计可满足ITER对其热工水力安全方面的要求。%Safety assessment on the dual-functional lithium lead test blanket module (DFLL-TBM) was performed with a modified version of RELAP5/MOD3 code in which the LiPb eutectic thermal-hydraulic sub-module was inserted .The DFLL-TBM and its ancillary cooling systems were modeled to conduct the computation and analysis for steady-state operation ,anticipated operational incidents and relevant accidents .Compu-tational results indicate that the maximum surface temperature of the first wall (FW) structural material is lower than the allowable value of 550 ℃ .For the accident analy-ses ,none of the pressure increases in ITER vacuum vessel (VV) ,port cell and TCWS vault induced by helium leaking is beyond the ITER safety limit of 0.2 MPa .No melting of the TBM box is found and the decay heat can be removed efficiently by the radiation heat transfer .With the current design ,DFLL-TBM can meet the thermal-hydraulic safety requirements from IT ER .

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

    International Nuclear Information System (INIS)

    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

  10. RELAP5-3D calculation of steam outlet header rupture of WWER-1000 NPP at hot zero power

    International Nuclear Information System (INIS)

    This presentation is devoted to one of the transients from the spectrum of steam line ruptures, which are analyzed for safety report purposes in Czech Republic. It is focused on the SGI steam outlet header rupture of WWER-1000 at hot zero power conditions analysis with advanced thermal-hydraulic code RELAP5-3D. The attention is addressed a reactor vessel and reactor core nodalization. The steam line rupture followed by steam release results in a strong drain oaf energy from primary circuit, what presents significant decreasing of primary coolant temperature and pressure. Then the feedback reactivity coefficients in connection with reduction of coolant temperature on the reactor inlet would cause introduction of positive reactivity and they could result in reactor restart (after its shutdown). Such transients are characterised by an unsymmetrical cool down of reactor and a strongly non-uniform distribution of power increase in the core. Safety analyses of such transients, where substantial changes oaf power distribution could occur, need apply thermal-hydraulic computational programs containing a 3D neutronic and thermal-hydraulic model of the reactor. The used computer codes, initial conditions for thermal-hydraulic calculations and basic results are presented. The DNBR value was monitored from the point of view of fuel integrity (Authors)

  11. Analysis of Air-Water Two Phase Flow for K-HERMES-HALF Experiment using RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Ha, Kwang Soon; Kim, Sang Baik; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Heo, Sun [KHNP Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The IVR (In-Vessel corium Retention) through the ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the integrity of the reactor pressure vessel during a severe accident in a nuclear power plant. This measure has been adopted in some low-power reactors such as the AP600, AP1000, and the Loviisa nuclear power plants as a design feature, and in the high-power reactor of the APR (Advanced Power Reactor) 1400 as an accident management strategy for severe accident mitigation. As part of a study on two-phase flow in the reactor cavity under external reactor vessel cooling in the APR1400, K-HERMES-HALF experiment (Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale) had performed at KAERI. This large-scale experiment using a half-height and half-sector model of the APR1400 uses the non-heating method of the air injection. In this research, K-HERMES-HALF test results had been evaluated by using RELAP5/MOD3 computer code to observe and evaluate the two-phase natural circulation phenomena through the annulus gap between the outer reactor vessel and the vessel insulation material

  12. Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Andrew M. Ward

    2011-01-01

    Full Text Available In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the GenPMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable.

  13. Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

    International Nuclear Information System (INIS)

    This report summarizes significant technical findings from the LP-FP-2 Experiment sponsored by OECD. It was the second, and final, fission product experiment conducted in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory. The overall technical objective of the test was to contribute to the understanding of fuel rod behavior, hydrogen generation, and fission product release, transport, and deposition during a V-sequence accident scenario that resulted in severe core damage. An 11 by 11 test bundle, comprised of 100 pre-pressurized fuel rods, 11 control rods, and 10 instrumented guide tubes, was surrounded by an insulating shroud and contained in a specially designed central fuel module, that was inserted into the LOFT reactor. The simulated transient was a V-sequence loss-of-coolant accident scenario featuring a pipe break in the low pressure injection system line attached to the hot leg of the LOFT broken loop piping. The transient was terminated by reflood of the reactor vessel when the outer wall shroud temperature reached 1517 K. With sustained fission power and heat from oxidation and metal-water reactions, elevated temperatures resulted in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. A description and evaluation of the major phenomena, based upon the response of on line instrumentation, analysis of fission product data, post-irradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented

  14. 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.

  15. Assessment study of the coupled code RELAP5/PARCS against the Peach Bottom BWR turbine trip test

    International Nuclear Information System (INIS)

    The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. In this framework, the Peach Bottom BWR turbine trip experiment 2 is considered. The test involves a rapid positive reactivity addition into the core generated by a water hammer load. To perform a numerical simulation of such phenomenon a reference case was calculated using the coupled code RELAP5/PARCS. An overall data comparison shows good agreement between calculated and measured pressure wave trend in the core region. However, the predicted power response during the excursion phase did not match correctly the experimental tendency. For this purpose, a series of sensitivity analyses have been carried out to identify the most probable reasons of such discrepancy. It was found out that the uncertainties related to the cross-sections modeling and to the thermal-hydraulic closure relationships are the main source of the incorrect power feedback response during the transient

  16. PSA support safety analysis using RELAP5 for the reactivity insertion event in 14 MW TRIGA reactor

    International Nuclear Information System (INIS)

    The paper presents the deterministic support analysis in case of Reactivity Insertion Accident (RIA) considered as initiating event in the PSA project. It studies the reactivity worth necessary to damage the research reactor fuel. In a previous PSA study the postulated initiating event due to a mistake in fuel handling was assumed as resulting from falling of one or at most two fuel bundles from the lifting device operated for core configuration rearrangements. This type of event was actually oc curing in the nineties. The paper gives some elements of the previous PSA model with respect to this initiating event, the event tree and the results of the accident aftermath. The focus of the paper is on the results of applying the thermalhydraulic code RELAP5 Mod 3.2 which uses a point kinetics model for studying the transient in case of different reactivity worths and insertion times. The results include evolutions of heat transfer mode, maximum temperature inside fuel elements and peak values of the power excursion. The conclusions highlight the possibility of infringement of the safety criteria for the TRIGA Ssr 14 MW reactor during the analyzed transients and also discuss the necessity of including this event in the PSA model. (authors)

  17. Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

    Science.gov (United States)

    Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

    2012-10-01

    Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

  18. Assessment study of RELAP5/MOD2 Cycle 36.04 based on pressurizer safety and relief valve tests

    International Nuclear Information System (INIS)

    This report presents a code assessment study based on full size relief and assisted safety valve (called SEBIM) tests performed on the CUMULUS valve test rig operated by Electricite de France (EdF). The increased awareness that the pressuriser safety and relief valves are not reliable under water blowdown conditions, has led to the design, testing and installation of so called assisted safety valves of which the SEBIM (TM) valves are an example. These valves, used in tandem, are gradually replacing the safety and relief valves on pressurisers in some European PWR's. Before installation at the plant, the Belgian safety authorities requested a thorough full scale testing of these valves on a test rig (CUMULUS) equipped with sufficient diagnostics to measure the characteristics of the valve. The Belgian architect-engineering firm TRACTEBEL was called upon the specify, order and test these valves for installation at the DOEL 1 and DOEL 2 power plants. These tests do provide sufficient data of high quality to justify an assessment study of the code RELAP-5 MOD-2 CYCLE 36 in the ICAP framework which is the subject of this report

  19. Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P; Jo, J H; Neymotin, L; Rohatgi, U S; Slovik, G

    1982-12-01

    This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented.

  20. Analysis of the IRIS pressurizer behavior in the presence of noncondensable gases using RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Eduarda da C.A.; Castrillo, Lazara S., E-mail: e.camedeiros@gmail.com, E-mail: lazara@poli.br [Universidade de Pernambuco, Recife, PE (Brazil). Escola Politecnica. Departamento de Engenharia Mecanica

    2015-07-01

    Insurge and outsurge phenomena are transient states and could be analyzed by thermodynamics principles, the pressurizer behavior will vary in response to mass flow changes. These surges can occur in the presence of noncondensable gases. On this paper, with the code RELAP5, the IRIS reactor pressurizer is described to analyze surges phenomena in their control volumes with non-condensable gases since they modify the pressure response. A set of three pipes components represents the pressurizer regions, connected with each other by single junctions components, the bottom volume control is connected to the primary circuit, represented by a time dependent volume component, through a time dependent junction component, which describes the mass flow behavior during surges through surge orifices. The hydrodynamic components representing the pressurizer are surrounded by heat structures, in addition there are heat structures inside the bottom volume control describing the behavior of electrical heaters, that operate in cases of outsurges. The analysis are intended to detail the behavior variables as pressure, temperature and volume of liquid inside the pressurizer during a water surge coming from the primary circuit or a water surge coming from the pressurizer to the primary circuit. (author)

  1. Implementation of a New DTSTEP Algorithm for use in RELAP5-3D and PVMEXEC Completion Report

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2010-12-01

    The PVM Coupling methodology for decomposing a complex model into domains onto which individual programs may be applied has proven effective for solving many multi-physics problems. There have been, from the outset, some detailed and/or long-running models that cause the process to fail. This project addressed the PVM coupling issues surrounding the DTSTEP subroutines on RELAP5-3D and PVMEXEC. Some 25 errors are listed in Tables 1 and 18 and in Section 11. These arise from deficiencies in the floating point calculation and testing of time steps, cumulative time, and time targets. The algorithmic replacement of floating point control of these items with integer based timestepping was developed and implemented. The result of the first phase, undertaken by the INL was that all but three of these issues were resolved. Moreover, two conceptual errors in DTSTEP that were not PVM coupling related were discovered and solved. The final, and most difficult three PVM Bettis User Problems, were solved during the Bettis phase of development and debugging. In 8 months since the conclusion of the project, no further DTSTEP related PVM Coupling errors have been reported.

  2. Post-test sensitivity analysis of OECD/CSNI ISP42 panda experiment by Relap5 code

    International Nuclear Information System (INIS)

    The present document deals with Relap5/Mod3.2 analysis of the International Standard Problem (ISP-42) exercise performed in PANDA facility on April 21-22, 1998. PANDA is installed at PSI (Paul Scherrer Institute). PANDA is a large-scale thermal-hydraulic test facility suitable for the simulation of passive containment for Advanced Light Water Reactors (ALWR). The work focuses phase A of the ISP-42 experiment, including the break in the main steam line, and the Passive Containment Cooling System Start-Up. The objective is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air and to observe the resulting system behavior. A detailed nodalization was set-up at the University of Pisa, in order to model 3-D flow paths with a 1-D code. The comparison between pre-test predictions and experimental data is discussed. Overall time behavior is reasonably well predicted, showing a rather good and robust overall code behavior in the simulation of the global test scenario. The results of a preliminary post-test analysis are discussed, including the comparison with the experimental data. (authors)

  3. RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

    International Nuclear Information System (INIS)

    Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)

  4. Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981

    International Nuclear Information System (INIS)

    This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented

  5. RGUI 1.0, New Graphical User Interface for RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    G. L. Mesina; J. Galbraith

    1999-04-01

    With the advent of three-dimensional modeling in nuclear safety analysis codes, the need has arisen for a new display methodology. Currently, analysts either sort through voluminous numerical displays of data at points in a region, or view color coded interpretations of the data on a two-dimensional rendition of the plant. RGUI 1.0 provides 3D capability for displaying data. The 3D isometric hydrodynamic image is built automatically from the input deck without additional input from the user. Standard view change features allow the user to focus on only the important data. Familiar features that are standard to the nuclear industry, such as run, interact, and monitor, are included. RGUI 1.0 reduces the difficulty of analyzing complex three-dimensional plants.

  6. RGUI 1.0, New Graphical User Interface for RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George Lee; Galbraith, James Andrew

    1999-04-01

    With the advent of three-dimensional modeling in nuclear safety analysis codes, the need has arisen for a new display methodology. Currently, analysts either sort through voluminous numerical displays of data at points in a region, or view color coded interpretations of the data on a two-dimensional rendition of the plant. RGUI 1.0 provides 3D capability for displaying data. The 3D isometric hydrodynamic image is built automatically from the input deck without additional input from the user. Standard view change features allow the user to focus on only the important data. Familiar features that are standard to the nuclear industry, such as run, interact, and monitor, are included. RGUI 1.0 reduces the difficulty of analyzing complex three dimensional plants.

  7. RELAP5-3D Developmental Assessment: Comparison of Versions 4.2.1i and 4.1.3i

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2014-06-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.2.1i and 4.1.3i. 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 changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  8. Benchmarking the RELAP5/MOD2.5 r-Θ model of an SRS [Savannah River Site] reactor to the 1989 L-Reactor tests

    International Nuclear Information System (INIS)

    Benchmarking calculations utilizing RELAP5/MOD2.5 with a detailed multi-dimensional r-θ model of the SRS L-Reactor will be presented. This benchmarking effort has provided much insight into the two-component two-phase behavior of the reactor under isothermal conditions with large quantities of air ingested from the moderator tank to the external loops. Initial benchmarking results have illuminated several model weaknesses which will be discussed in conjunction with proposed modeling changes. The benchmarking work is being performed to provide a fully qualified RELAP5 model for use in computing the system response to a double ended large break LOCA. 5 refs., 14 figs

  9. Comparison of an integral response scaling method with Ishii's scaling method and its validation using RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    An integral response scaling method for a reduced-height test facility is suggested and the scaling laws derived from it are compared with Ishii's scaling. In the present scaling method it turns out that flow velocities in the vertical channel and through the break area or injection area should be preserved. RELAP5/MOD3.2 code calculations of pot-boiling, blowdown, heat transfer in Steam Generator(SG) and off-take are conducted for the validation of the present scaling method. Four scaled-down models are designed based on the present method and Ishii's scaling method given length and area scales of 1/5 and 1/100, respectively. RELAP5/MOD3.2 calculations show that the scaled-down model based on the present scaling method well maintains the similarity of the nondimensional mixture level in pot-boiling, the nondimensional pressure in blowdown and the heat transfer coefficient in SG

  10. RELAP5-3D Developmental Assessment: Comparison of Versions 4.3.4i and 4.2.1i

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.3.4i and 4.2.1i. 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 changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  11. A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.Th. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

    1995-09-01

    As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

  12. RELAP5/MOD3.2 Analysis of a VVER-1000 Reactor with UO2 Fuel and Mixed-Oxide Fuel

    International Nuclear Information System (INIS)

    A RELAP5/MOD3.2 model of the VVER-1000/MODEL V320 nuclear power plant was modified and a large-break loss-of-coolant accident (LBLOCA) in the cold leg was simulated. In this analysis, the core consisted of 162 UO2 assemblies and 1 mixed-oxide assembly. The results from the simulation were compared with the results from a similar study performed with the Russian computer program TECH-M. An uncertainty analysis was performed on the peak cladding temperature following a similar methodology called code scaling, applicability, and uncertainty. Monte Carlo calculations were performed using the response surface inferred from 15 runs of RELAP5 calculations. The result of this study shows that the emergency core coolant system would be sufficient to keep the cladding temperature during the LBLOCA scenario well below the required maximum limit

  13. Restructuring the electronic medical record to incorporate full digital signature capability.

    OpenAIRE

    Zuckerman, A. E.

    2001-01-01

    The security of Electronic Medical Records can be enhanced by the addition of digital signatures that guarantee data integrity, authenticate the signer, and establish non-repudiation through the use of public key encryption. The task is complicated by the contribution of multiple providers to an encounter and the entry of data at multiple points in time Dividing encounters into an episode of care and redesigning the data model of the EMR will facilitate full signature capabilities. Generation...

  14. SPES-99 IBLOCA analysis with the RELAP5 Mod3.2 code

    International Nuclear Information System (INIS)

    SPES is an experimental facility which allows the assessment of thermalhydraulic codes through the simulation of a wide range of physical phenomena characterising different accident scenarios in PWRs. In 1999, in view of a possible proposal for an international cooperative programme, ENEA provided the SIET company with limited funding to restore the facility after five-year shutdown and ordered SIET a ''demonstration'' experiment where the new SPES configuration and the test conditions were selected to exploit in a valuable way the unique and outstanding characteristics of the facility. The test, defined by a working group composed of ENEA, ANPA, JRC Ispra, ANSALDO, Pisa University and SIET, consisted of a 10'' equivalent break in Cold Leg, starting from full power and full pressure conditions. This document deals with the analysis of the comparison between calculated results and experimental data. Moreover, it underlines the SPES facility capability to simulate IB-LOCAs in addition to SB-LOCAs and other kind of transients for which it was designed. (author)

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

    OpenAIRE

    Siniša Šadek; Srđan Špalj; Bruno Glaser

    2010-01-01

    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 ...

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

    International Nuclear Information System (INIS)

    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

  17. Assessment of RELAP5/MOD3/V5M5 against the UPTF Test No. 11 (countercurrent flow in PWR hot leg)

    International Nuclear Information System (INIS)

    Analysis of the UPTF Test No. 11 using the open-quotes best-estimateclose quotes computer code RELAP5/MOD3/Version 5M5 is presented. Test No. 11 was a quasi-steady state, separate effect test designed to investigate the conditions for countercurrent flow of steam and saturated water in the hot leg of a PWR. Without using the code's new countercurrent flow limitation (CCFL) model, RELAP5/MOD3/V5M5 overestimated the mass flow rate of back down flowing water up to 35% (1.5 MPa runs) and 43% (0.3 MPa runs). This is the most obvious difference to RELAP5/MOD2, which did not allow enough countercurrent flow. From the point of view of performing plant calculations this is certainty an improvement, because the new junction-based CCFL option could be used to restrict the flows to a flooding curve defined by a user-supplied correlation. Very good agreement with the experimental data for 1.5 MPa -- which are relevant for SBLOCA reflux condensation conditions -- could be obtained using the code's new CCFL option in the middle of the inclined part (riser) of the hot leg. Using the same CCFL correlation for the simulation of 0.3 MPa test series -- typical for reflood conditions -- the code underestimated by 44% the steam mass flow rate at which complete liquid carry over occurs. An unphysical result was received using a CCFL correlation of the Wallis type with the intercept C = 0.644 and the slope m = 0.8. The unphysical prediction is an indication of possible programming errors in the CCFL model of the RELAP5/MOD3/V5M5 computer code

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. V1000CT-1 benchmark analyses with the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems

    International Nuclear Information System (INIS)

    Full text of publication follows:Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and 3 of 4 MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Control rods were not changing their positions during the transient. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical nodalization schemes, MCP characteristics, boundary conditions and the benchmark-specified nuclear data library. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermohydraulic models of the system codes RELAP5 and ATHLET. (authors)

  1. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET

  2. Comparisons of steady-state and transient thermal hydraulic results from SAS-DIF3DK and RELAP5 mod 3.2 for an RBMK reactor

    International Nuclear Information System (INIS)

    The SAS-DIF3DK code couples a detailed 3-D neutron kinetics treatment with a detailed thermal hydraulics treatment. One goal of the work on SAS-DIF3DK is to produce a detailed code that will run a wide range of transients in real time. Achieving this goal will require efficient numerical methods and efficient coding, and it will probably require the use of multiple processors for larger problems. In order to obtain clean code-to-code thermal hydraulics comparisons with a recognized and established code, a detailed thermal hydraulic model was set up for an RBMK assembly and its associated piping, The same identical input model was implemented in both SAS-DIF3DK and RELAP5 mod 3.2. Both steady-state and transient thermal hydraulics calculations were made with this model. Except for cladding temperatures in one transient, the SAS-DIF3DK results were similar to or almost identical to the RELAP5 results, and SAS-DIF3DK ran an order of magnitude or more faster than RELAP5. The cladding temperature differences can be explained in terms of different post-DNB models and heat transfer coefficients

  3. RELAP5/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    RELAP5/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients. In developing the 4-loop plant model, a three loop W plant model was modified at ANL into a four loop plant. This plant model retained the two loop nodalization of the original model, but split them such that one loop represented the faulted steam generator (broken at the outlet just past the integral flow restrictor) while the other loop represented all three of the intact loops. In the case of the 3-loop plant model, a RELAP5 model was modified to perform this replication calculation. One loop represented the faulted steam generator (broken between the steam generator and the non-integral flow restrictor) while the other loop represented the remaining two intact loops. In both cases (3-loop and 4-loop), the reactor vessel was modelled as two parallel channels to accommodate the Westinghouse steam line break methodology

  4. An analysis of ROSA-IV/LSTF 10% main steam line break test run SB-SL-01 using RELAP5/MOD3

    International Nuclear Information System (INIS)

    This paper presents RELAP5/MOD3 code calculations of a 10% main steam line break test, designated as RUN SB-SL-01, conducted using the ROSA-4 Large Scale Test Facility (LSTF). The RELAP5/MOD3 input deck of LSTF, which includes 189 volumes, 200 junctions, and 180 heat slabs, was modeled to obtain best-estimate predictions of several important features during the main steam line break accident in order to property evaluate the consequences of this accident. The main conclusions drawn were that the results of RELAP5/MOD3 code calculations were in reasonable agreements with test RUN SB-SL-01, especially for the trends of key parameters. Detailed investigations indicated minor discrepancies in RCS pressure during the period of time that voiding occurred in the upper head. This is possible due to emptying of the pressurizer and voiding in the upper head. Sensitivity studies were also performed for the break junction discharge coefficient and the separator drain line loss coefficient. These parameters had significant effects on the steam quality on the secondary side and on the break flow through the change of water inventory on the secondary side. This phase separation process was adequately predicted during all transients with break junction discharge coefficient of 0.85 and separator drain line loss coefficient of 10

  5. RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlova@inrne.bas.bg; Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, A.E. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Gencheva, R.V. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg

    2006-02-15

    This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit. RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters. This paper discusses the results of the thermal-hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences. This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP)

  6. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  7. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

    International Nuclear Information System (INIS)

    This paper presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is in an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  8. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

    International Nuclear Information System (INIS)

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  9. Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

    International Nuclear Information System (INIS)

    This report records the results of the code validation for the one-dimensional module of the MARS-KS 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 Developmental 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 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 result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  10. Transient analysis of mercury experimental loop using the RELAP5 code. 3rd report. Transient analysis using mercury properties

    International Nuclear Information System (INIS)

    In order to promote the Neutron Science Project of JAERI, the design of a 5MW-spallation target system is in progress with the purpose of producing a practical neutron application while at the same time adhering to the highest levels of safety. To establish the safety of the target system, it is important to understand the transient behaviors during anticipated operational events of the system, and to design the safety protection systems for the safe termination of the transients. This report presents the analytical results of transient behaviors in the mercury experimental loop using mercury properties. At first, the analytical pressure distributions were compared with experimental data measured with the mercury experimental loop. The modeling data were modified to reproduce the actual pressure distributions of the mercury experimental loop. Then a loss of forced convection and a loss of coolant accident were analyzed. In the case of the pump trip, the transient analysis was conducted using two types of mercury pumps, the mechanical type pump with moment of inertia, and the electrical-magnetic type pump without moment of inertia. The results show there was no clear difference in the two analyses, since the mercury had a large inertia, which was 13.5 times that of the water. Moreover, in the case of a pipe rupture at the pump exit, a moderate pressure decrease was confirmed when a small breakage area existed in which the coolant flowed out gradually. Based on these results, it was appeared that the transient fluctuation of pressure in the mercury loop would not become large and accidents would have to be detected by small fluctuations in pressure. Based on these analyses, we plan to conduct a simulation test to verify the RELAP5 code, and then the analysis of a full-scale mercury system will be performed. (author)

  11. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E.J.; Chung, B.D.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1993-06-01

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator.

  12. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    International Nuclear Information System (INIS)

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator

  13. Analysis of thermohydraulic limits during WWER-1000 nuclear power plants heat-up using RELAP5 system code

    International Nuclear Information System (INIS)

    Plant heat-up is a process which all operating systems such as primary coolant circuit, pressurizer, primary and secondary sides of the steam generators and etc. are transferred from a cold shutdown to a hot standby status. During plant heat-up, some thermohydraulic limitations such as maximum and minimum allowable pressure and maximum rate of increase in pressure and temperature which are recommended by plant commissioning program and NPP safety related documents should be considered. Maximum allowable pressure prevents brittle fracture in reactor vessel, Minimum allowable pressure in the inlet of the reactor coolant pumps (RCPs) prevent pump cavitations and maximum allowable rate of increase in temperature and pressure respectively prevent thermal and mechanical shocks. Thus, tuning pressure and temperature increasing rates during plant heat-up is important from plant safety point of view. The RELAP5 system code was used to model and analysis the behavior of WWER-1000 plants during heat-up. In plant heat-up, at first the primary circuit pressure increases by injection of N2 gas into pressurizer in order to provide minimum required NPSH (net positive suction head) for operation of the RCPs. After short time RCPs are turned on to operate which increases the primary coolant circuit temperature through friction losses. At a time which is specified by heat-up procedure the pressurizer heaters are turning on to increase the primary circuit pressure. Heat transfer from primary to secondary side in the steam generators causes increasing of the secondary side temperature and pressure. Temperature and pressure of primary and secondary circuits increase until plant reaches to hot standby condition. The results show that the thermohydraulic parameters during plant heat-up are in an acceptable range and have a good agreement with available data in technical documents. (authors)

  14. Analysis of experiments for steam condensation in the presence of noncondensable gases using the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    A computational investigation of experiments involving the condensation phenomenon in the presence of noncondensable gases was performed. The RELAP5/MOD3 thermal-hydraulic code was utilized for this analysis. Two separate-effects experiments were studied, which are relevant to actual situations encountered in the industry. The first experiment involved condensation of steam in an inverted U-tube when nitrogen is present. A constant flow of steam was injected into the U-tube and condensed along its surface. The condensing length was a function of the injected nitrogen rate and the secondary temperature. The code predicted an active condensation zone with unimpeded heat transfer and a passive zone with no heat transfer. The lengths of these zones agree with the experimental data. The gas temperatures in the U-tube were favorably predicted except for a discrepancy where the calculated primary temperatures were lower than the secondary temperatures for several cases. Active nitrogen contents in the tube were underpredicted by the code. The second experiment investigated was the Massachusetts Institute of Technology's steam condensation experiment. This experiment modeled the proposed containment cooling system for advanced reactors. Steam was generated in a vessel in which air was present. The steam in the steam-air mixture condensed on the surface of a cooled copper cylinder. Computational predictions of this experiment revealed that heat transfer coefficients vary with air fraction. Calculated heat transfer coefficients were compared with the data, and it was found that the results were better for higher system pressures than for lower pressures

  15. Restructuring the electronic medical record to incorporate full digital signature capability.

    Science.gov (United States)

    Zuckerman, A E

    2001-01-01

    The security of Electronic Medical Records can be enhanced by the addition of digital signatures that guarantee data integrity, authenticate the signer, and establish non-repudiation through the use of public key encryption. The task is complicated by the contribution of multiple providers to an encounter and the entry of data at multiple points in time Dividing encounters into an episode of care and redesigning the data model of the EMR will facilitate full signature capabilities. Generation of digital signatures is best accomplished using microprocessors on smart cards that control visibility of the private keys and assist in user authentication. The Java Programming Language including cryptography extensions and a smart card API is a useful tool for adding digital signature to an EMR. Inter-operability of signatures and continuity of signature will require attention to standards and preservation of cryptography and authentication certificate archives. Digital signatures will need to accommodate changes in data storage formats when information is transported between EMR systems using XML or other transaction standards because the original signatures will not validate if the data storage format changes. The costs of adding digital signature to EMR mandates serious examination of the business case for digital signature within an EMR as compared with transactions such as electronic prescriptions. At present, there is no regulatory requirement for digital signature of an EMR. PMID:11825294

  16. RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究%Applicability Research of RELAP5/MOD3.3 for Small Break Loss of Coolant Accident of NPP With Passive Safety System

    Institute of Scientific and Technical Information of China (English)

    徐财红; 史国宝

    2014-01-01

    AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA 研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的REL A P5/M OD3.3的适用性,以A P1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的REL A P5/M OD3.3的计算结果与试验数据符合较好。%The passive core cooling system is used in AP 1000 to mitigate the small break loss of coolant accident (SBLOCA) .The RELAP5/MOD3.3 code is generally applicable to the traditional NPP SBLOCA research , but for the passive NPP SBLOCA , its applicability will need further study and evaluation . Based on the analysis of the important phenomenon of the SBLOCA of the passive NPP , the RELAP5/MOD3.3 code was assessed and modified . In order to verify the applicability of the modified RELAP5/MOD3.3 code ,the DBA-02 and NRC-05 cases of APEX-1000 which was the test facility for verifying AP1000 small break loss of coolant accident ,were simulated . It shows good agreement between the results of the modified RELAP5/MOD3.3 code and experiment data .

  17. 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)

    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.)

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

    International Nuclear Information System (INIS)

    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.

  19. MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code

    International Nuclear Information System (INIS)

    For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO2 fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)

  20. Coupled RELAP5/PARCS main steam line break calculations before and after a power up-rate of a Pressurized Water Reactor

    International Nuclear Information System (INIS)

    This paper reports on the analysis of a hypothetical Main Steam Line Break (MSLB) at the Swedish Ringhals-3 Pressurized Water Reactor. Ringhals-3 is of a three-loop Westinghouse design with an original design power of 2783 MWth. A power up-rate to 3142 MWth is planned for this reactor, and the consequences of this up-rate on different MSLB scenarios are investigated in this paper. The codes used in the analysis are PARCS and RELAP5. The PARCS model accounts for the full heterogeneity of the core, with explicit modelling of the top, bottom, and radial reflectors. The RELAP5 model accounts for each individual fuel assembly in the core, which means that there is a radial one-to-one correspondence between the thermal-hydraulic and neutronic models. Both the stand-alone PARCS model and the stand-alone RELAP5 model were earlier validated against steady-state measurements at the Ringhals-3 unit. Validations for the coupled model using measured transients that occurred at Ringhals-3 were also successfully carried out in the past. In the MSLB simulations reported in this paper, a number of different cases are considered, where the influence of core burnup, of flow mixing in the lower plenum, and of the power level at Hot Full Power conditions is investigated. Simulations performed at Hot Zero Power conditions demonstrate the time between the reactor scram and the break has a strong influence on the behaviour of the system (effect due to the delayed neutrons). For longer shutdown times, a return to criticality is possible. The simulations also showed that the system returns to criticality if the safety injection fails. (authors)

  1. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    OpenAIRE

    Takeshi Takeda

    2014-01-01

    RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM) measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW) injection into the secondary-side of both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depre...

  2. Perfeccionamiento del modelo de vasija del reactor de la central de Ascó para el código de cálculo Relap5

    OpenAIRE

    Berggren Durall, Alberto

    2012-01-01

    El objetivo principal del proyecto realizado es el de perfeccionar el modelo de vasija de la central nuclear de Ascó utilizando el código de cálculo RELAP5/mod3.3. Con este fin se proponen y estudian tres variaciones en la modelación de la vasija perteneciente al modelo general de planta y se escogen dos escenarios relevantes para analizar y comparar los comportamientos. En primer lugar, a partir de un cambio real efectuado en la CNA se desarrolla una nueva nodalización de l...

  3. Validation of six-loop RELAP5 model of VVER-440/V213 unit with transients measured in the Paks NPP

    International Nuclear Information System (INIS)

    In the AGNES project a six-loop input model of then third unit of the Paks NPP has been developed for the RELAP5/mod2 computer code. To verify that model, transients measured in Paks NPP were chosen for simulation: trip of main circulating pump and scram after a spurious signal from the safety system. The analyses of the transients show that, with only a very few modifications in the input model, the results are in a good agreement with the measured data. (author)

  4. Application Of VISA And RELAP5 Software To The Modelling, Simulation And Calculation Of The Thermal Hydraulic System Of Nuclear Power Plans With VVER-Type Reactors

    International Nuclear Information System (INIS)

    This paper presents an overview of the VISA (Visual System Analyzer) and RELAP5 (Reactor Excursion and Leak Analysis Program, version 5) softwares. Preliminary results of modelling, simulation and calculation of the thermal hydraulic system of a practical VVER-type PWR nuclear power plan (NPP) using the software are presented as well. This research has important practical implications, especially for the nuclear power plans that are being built in Ninh Thuan, Vietnam. The reactors adopted in the Ninh Thuan NPP would be the Russian VVER PWR. At this point, our research is a basic important step towards a practical case study. (author)

  5. Vectorization, parallelization and implementation of nuclear codes [MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3] on the VPP500 computer system. Progress report 1995 fiscal year

    International Nuclear Information System (INIS)

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  6. Assessment of steam condensation model with the presence of non-condensable gas in a vertical tube using RELAP5 Mod 3.2 code and MIT exp. Data

    International Nuclear Information System (INIS)

    The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of AP1000, APR1400, AES-2006, the Passive Containment Cooling System (PCCS) of AP1000 design, and Isolation Condensation System (ICS) of ESBWR design. The accumulation of the non-condensable gas inside the condensing tubes can significantly reduce the level of heat transfer which affects the heat removal capacity in accident condition and impacts plant safety. The objective of the present work is to assess the analysis capability of two wall film condensation models of RELAP5/Mod3.2 with the presence of non-condensable gas in a vertical tube on condensation experiments performed at MIT, USA. The results of the simulations and experimental data show the similar tendencies that the heat transfer coefficients increase as the inlet steam-non condensable gas mixture flow rate increases, the inlet steam-non-condensable gas mass fraction decrease, and the inlet saturated steam temperature decrease. (author)

  7. Simulation with RELAP5/MOD3.3 of a postulated 10% hot leg break in Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    This paper presents the simulation results of a 10% break in the hot leg of Angra 2 nuclear power plant, which was run with the computer code RELAP5/MOD3.3. The initial steady state conditions for this simulation are in agreement with the experiment named SB-HL-02 that was conducted in the Large Scale Test Facility in the Rig of Safety Assessment-IV program (ROSA-IV/LSTF). The main boundary conditions specified for the simulation were: high pressure injection system (HPI) and auxiliary feedwater system (AFW) were assumed to be unavailable; and loss of offsite power was assumed to occur concurrently with scram. The results obtained were scaled down and compared with the ROSA-IV/LSTF test, which was performed with the same boundary conditions. This activity was executed in the scope of IAEA research project (CRP J72005) - Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3.3 Code Applying CIAU Methodology. (author)

  8. 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)

    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

  9. RELAP5-3D Modeling of Heat Transfer Components (Intermediate Heat Exchanger and Helical-Coil Steam Generator) for NGNP Application

    International Nuclear Information System (INIS)

    The Next Generation Nuclear Plant project is aimed at the research and development of a helium-cooled high-temperature gas reactor that could generate both electricity and process heat for the production of hydrogen. The heat from the high-temperature primary loop must be transferred via an intermediate heat exchanger to a secondary loop. Using RELAP5-3D, a model was developed for two of the heat exchanger options a printed-circuit heat exchanger and a helical-coil steam generator. The RELAP5-3D models were used to simulate an exponential decrease in pressure over a 20 second period. The results of this loss of coolant analysis indicate that heat is initially transferred from the primary loop to the secondary loop, but after the decrease in pressure in the primary loop the heat is transferred from the secondary loop to the primary loop. A high-temperature gas reactor model should be developed and connected to the heat transfer component to simulate other transients

  10. Investigation of Local Effects Influence on Results of Design Basis Accident Analysis of WWER-440 Reactor Using RELAP5-3D Code

    International Nuclear Information System (INIS)

    One of the most important tasks in today's nuclear power plant safety analysis is a simulation of physical processes at nuclear facilities which accounts for 3-dimensional effects in the core and downcomer of reactor. System coupled thermo-hydraulic/neutron-kinetic code RELAP5-3D, which is a modeling tool provided to University of Kyiv by US DOE in a frame of International Nuclear Safety Program, allows simulation of variable in time spatial distribution of neutron flux in a core and also includes special components for 3D modeling of thermo-hydraulics. A model of Rivne NPP Unit 1 with WWER-440/V-213 type reactor has been developed for RELAP5-3D code. A scenario of 'Main steam line break' design basis accident has been calculated using this model. Such a problem can be characterized by intensive overcooling of a primary coolant in affected loop and, taking into account partial mixing of coolant from different primary loops, a non-uniform cooling of reactor core. Obtained results have been compared with the results obtained by model, which has been used at Design Based Accidents analysis, performed at specified unit.(author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.T. [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, 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.

  12. Conversion of control systems, protection and engineering safeguard system signals of Almaraz NPP model from RELAP5 into TRAC-M

    International Nuclear Information System (INIS)

    In the scope of a joint project between the Spanish Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) about the USNRC state-of-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As a part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. One of the goals of the project is to analyze the conversion of control blocks, signal variables and trips in order to correct modelling all instrumentation and control systems, and also protection and engineering safeguard system-signals of the NPP. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper describes the problems found in the conversion and the solutions achieved.(author)

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

    International Nuclear Information System (INIS)

    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)

  14. Rod ejection accident 3D-dynamic analysis in Trillo NPP with RELAP5/PARCS V2.7 coupled codes

    International Nuclear Information System (INIS)

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Trillo NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCSv2.7. In this work, we present the results of the REA analysis at 30% of the rated power at BOC. In the thermalhydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 177 x 32 active nodes, considering 28 different fuel elements with 867 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  15. Qualification of the TH-3DNK coupled code RELAP5/PARCSV2.7 against a BWR unstable point at peach bottom nuclear power plant

    International Nuclear Information System (INIS)

    In this work, three dimensional time domain BWR stability analysis were performed on a new analysis point (PTUPV), which is inside the exclusion region with a core mass flow of 4660.1 kg/s (34% of the core rated mass flow) and total reactor power of 1997.8 MW (60.7 of the core rated reactor power), using the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by the control rod movement as it is usual performed in Nuclear Power Plants. The transient starts with the control rod movement shown. The control rods move in 6 seconds; at the end of the movement the majority of the banks are completely withdrawn and only the bank 7 is almost completely inserted. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. With the aim to make a more careful analysis of the instability recognized in the RELAP5/PARCS simulation, using the nuclear cross-sections provided by the transient calculation performed with the coupled codes, a number of analyses with the nodal modal code VALKIN has also been carried out.

  16. 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.

  17. RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2012-01-01

    Full Text Available A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

  18. Rod ejection accident 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS V2.7 coupled codes

    International Nuclear Information System (INIS)

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Almaraz NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCS v2.7. In this work, we present the results of the REA analysis at hot zero power at BOC with all control rods inserted. In the thermal-hydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 157 x 24 active nodes, considering 13 different fuel elements with 291 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  19. 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)

    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

  20. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Ronaldo Celem

    2001-10-15

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  1. Estudi comparatiu de les opcions de modelització del combustible nuclear disponibles en els codis RELAP5 i SCDAP. Aplicació a transitoris base de disseny

    OpenAIRE

    Barberà Parramon, Miquel

    2012-01-01

    El comportament d'una central nuclear en escenaris accidentals constitueix l'objecte d'aquest projecte. Aquests comportaments transitoris s'estudien normalment mitjançant l'ús de codis de càlcul i models. El principal objectiu del projecte és determinar la conveniència d'utilitzar mètodes elaborats i complexes en detriment de mètodes més simples i ràpids en l'estudi de transitoris base de disseny mitjançant l'ús del codi de càlcul RELAP5 / SCDAP3.5. Amb aquesta finalitat es proposen d...

  2. Adjoint Sensitivity Analysis of the RELAP5/MOD3.2 Two-Fluid Thermal-Hydraulic Code System - II: Applications

    International Nuclear Information System (INIS)

    This work presents results that illustrate the validation of the Adjoint Sensitivity Model (ASM-REL/TF) corresponding to the two-fluid model with noncondensable(s) used in RELAP5/MOD3.2. This validation has been carried out by using sample problems involving (a) a liquid phase only, (b) a gas phase only, and (c) a two-phase mixture (of water and steam). Thus, the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability have been verified of the numerical solution of the ASM-REL/TF when only the gas phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid- and two-phase fluids, respectively, were replaced with pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations.This work also illustrates the role that sensitivities of the thermodynamic properties of water play for sensitivity analysis of thermal-hydraulic codes for light water reactors. The well-known 1993 ASME Steam Tables are used to present typical analytical and numerical results for sensitivities of the thermodynamic properties of water to the numerical parameters that appear in the mathematical formulation of these properties. Particularly highlighted are the very large sensitivities displayed by the specific isobaric fluid and gas heat capacities Cpf and Cpg, respectively; the specific

  3. RELAP5/MOD3 predictions of the effects of containment back pressure on long-term cooling for the AP600 design

    International Nuclear Information System (INIS)

    The RELAP5/MOD3 code has been used to perform calculations of a 2 inch cold leg break in the Westinghouse AP600 reactor design at two different containment back pressures for the purpose of identifying back pressure effects on the unique safety systems performance. Containment back pressure was held constant at 1.01x105 Pa (14.7 psia) for one case and at 3.10x105 Pa (45.0 psia) for the other case. The predictions show little difference between the two cases up to the time when the 4th stage ADS flow becomes unchoked. Intermittent and continuous IRWST injection resulted from the low back pressure case and the high back pressure case, respectively

  4. The addition of non-condensable gases into RELAP5-3D for analysis of high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of- coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  5. The application of Cathare 1 V1.3 to LOBI small break Loca experiments and a comparison with RELAP5/MOD2

    International Nuclear Information System (INIS)

    The paper presents an overview of the application of CATHARE V1.3 to LOBI Small Break LOCA tests, performed at Dipartimento di Costruzioni Meccaniche e Nucleari of Pisa University. In particular, the development of a new nodalization of LOBI facility is discussed along with the analysis of tests A2-81 (1% CL break). A1-83 (10% CL break) and A1-84 (10% HL break). In the second part of the paper, uncertainties are outlined which are typical of the analysis of experiments in integral test facilities. Finally, on the basis of the application of RELAP5/MOD2 to the analysis of test A2-81, a judgement is given about the behaviour of the two codes emphasizing the related advantages and disadvantages

  6. Modelado de un ciclo de potencia de CO2 supercrítico para reactores de fusión utilizando RELAP5-3D

    OpenAIRE

    Batet Miracle, Lluís; Álvarez Fernández, Josep Maria; Mas de les Valls Ortiz, Elisabet; Pérez, Marina; Martínez Quiroga, Víctor; Reventós Puigjaner, Francesc Josep; Sedano Miguel, Luis Angel

    2013-01-01

    En el marco del programa español de Tecnología de Fusión TECNO_FUS se ha avanzado en la definición de sistemas para DEMO, entre ellos las unidades reproductoras de tritio y el ciclo de potencia. Para las primeras, se ha propuesto un diseño modular a doble refrigerante (PbLi-He). Para la conversión de potencia térmica a eléctrica se han investigado ciclos de CO2 supercrítico. Mediante el código de sistema RELAP5-3D© se ha simulado un ciclo de potencia de CO2 supercrítico con recompresión. E...

  7. Analisis de un transitorio de inyección de Boro en un reactor PWR con el código acoplado RELAP5/PARCSv2.7

    OpenAIRE

    Garcia-Fenoll, Marina; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael; Verdú Martín, Gumersindo Jesús

    2011-01-01

    En este trabajo se presenta la implementación de una nueva prestación en el código acoplado RELAP5/PARCS v2.7 que permite analizar transitorios en los que se produce una variación de a concentración de boro en el núcleo. La implementación de la opción de inyección/dilución de boro consiste en la modificación del código fuente para que sea capaz de utilizar tablas de secciones eficaces distintas para diversas concentraciones de boro e interpolar entre ellas, así como en la mejora de la informa...

  8. Post-test analysis with RELAP5/MOD2 of ROSA-IV/LSTF natural circulation test ST-NC-02

    International Nuclear Information System (INIS)

    Results of post-test analysis for the ROSA-IV/LSTF natural circulation experiment ST-NC-02 are presented. The experiment consisted of many steady-state stages registered for different primary inventories. The calculation was done with RELAP5/MOD2 CYCLE 36.00. Discrepancies between the calculation and the experiment are observed: the core flow rate is overestimated at inventories between 80 % and 95 %; the inventory at which dryout occurs in the core is also much overestimated. The causes of these discrepancies are studies through sensitivity calculations and the following key parameters are pointed out: the interfacial friction and the form loss coefficients in the vessel riser, the SG U-tube multidimensional behaviour, the interfacial friction in the SG inlet plenum and in the pipe located underneath. (author)

  9. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  10. Assessment of MIT and UCB wall condensation tests and of the pre-release RELAP5/MOD3.2 code condensation models

    International Nuclear Information System (INIS)

    In recent years, a new class of reactor designs has been proposed that utilize passive safety systems. General Electric has developed a Simplified Boiling Water Reactor (SBWR) design that relies on such passive systems. The SBWR has two passive cooling systems that involve energy transfer by condensation. These are the isolation condenser system (ICS) and the passive containment cooling systems (PCCS). It is important that such heat transfer phenomena be correctly understood and quantified. The General Electric Company has sponsored tests at the Massachusetts Institute of Technology (MIT) and at the University of California at Berkeley (UCB) to obtain data simulating PCCS conditions. Data was obtained with pure steam, steam-air mixtures and steam-helium mixtures. INEL has been contracted by the NRC to evaluate these tests and assess existing condensation heat transfer correlations against the test data. This report assesses the relevance of the tests to SBWR conditions and shows RELAP5/MOD3.2 predictions of the tests

  11. Assessment of full power turbine trip start-up test for C. Trillo 1 with RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M. [Consejo de Seguridad Nuclear, Madrid (Spain)

    1993-07-01

    C. Trillo I has developed a model of the plant with RELAP5/MOD2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other PWRs of different design, the main difference being that the reactor is not tripped: a reduction in primary power is carried out instead. Pre-test calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitation Systems, were identified and post-test calculations had been carried out. The results show a good agreement with data for all the compared variables.

  12. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  13. Simulation of steam condensation in the presence of noncondensable gases in horizontal condenser tubes using RELAP5 for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as residual heat removal (RHRS) and passive containment cooling system (PCCS). Condensation studies of steam and noncondensable gases mixtures in these heat exchangers are very important due to the phenomena multidimensional nature and the condensate stratification effects. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and noncondensable gases (air) inlet mass fractions. The test section is three meters long and consists of two concentric tubes containing pressure, temperature and flow rate sensors. The internal tube, called condenser, contains steam-air mixture flow and external tube is a counter current cooler with water flow rate at low temperature. This test section was modeled and simulations were performed with RELAP5 code. Experimental tests were carried out for 200 to 400 kPa inlet pressure and 5, 10, 15 and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 and 400 kPa pressure conditions and showed good agreement. However, for 400 kPa inlet steam pressure and inlet air mass fractions above 5%, the simulated temperatures are lower than the experimental data at the final third from the inlet condenser tube, indicating a code overestimation of heat transfer coefficient. New correlations for heat transfer coefficient in these steam-air conditions must be theoretical and experimentally studied and implemented in RELAP5 code for better representing the condensation phenomena. (author)

  14. RELAP5/MOD3.3 analysis of the Reactor Coolant Pump Trip event at NPP Krsko for different transient scenarios

    International Nuclear Information System (INIS)

    In the paper the results of RELAP5/MOD3.3 analysis of Reactor Coolant Pump (RCP) Trip event at NPP Krsko for different transient scenarios are presented. The RCP Trip event occurred at NPP Krsko on 25.02.2002 at night when the shift crew began to reduce the power of the plant because of the increase of the temperature reading of the upper radial bearing of the RCP 2. After about 50 minutes when the reactor power was reduced to 28 %, the operators shut down the reactor and the RCP 2 since the temperature reading was still high. Following reactor trip an unexpected steam leak has been identified and the Main Steam Isolation Valves (MSIVs) isolation was performed. According to the plant event report the Steam Generator (SG) Power Operated (PORV) valve did not open at the setpoint. The SG 1 safety valve (SV) 1 opened twice. For the second time the SG 1 SV 1 opened at lower pressure than the nominal setpoint. The motor driven Auxiliary Feedwater (AFW) pumps were stopped due to overheating of the axial bearings and the turbine driven AFW pump was started. After stabilizing the steam line pressure below the SG 1 safety valve opening setpoint and with leaking path on the secondary side isolated, the MSIVs were opened. In the paper different transient scenarios and operator actions were analyzed using RELAP5/MOD3.3 computer code in order to support the resolution of potential safety issues identified as a result of an event and to analyze the severity of consequences in the event of additional failures. (author)

  15. RELAP5/MOD3.2 Sensitivity Analysis Using OECD/NEA ROSA-2 Project 17% Cold Leg Intermediate-break LOCA Test Data

    International Nuclear Information System (INIS)

    An experiment simulating a PWR intermediate-break loss-of-coolant accident (IBLOCA) with 17% break at cold leg was conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the experiment, core dryout took place due to rapid drop in the core liquid level before loop seal clearing (LSC). Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to counter-current flow limiting (CCFL) by high velocity vapor flow, causing further decrease in the core liquid level. The post-test analysis by RELAP5/MOD3.2.1.2 code revealed that cladding surface temperature of simulated fuel rods was under-predicted due to later major core uncovery than in the experiment. Key phenomena and related important parameters, which may affect the core liquid level behavior and thus the cladding surface temperature, were selected based on the LSTF test data analysis and post-test analysis results. The post-test analysis conditions were considered as 'Base Case', for sensitivity analysis to study the causes of uncertainty in best estimate methodology. The RELAP5 sensitivity analysis was performed by changing the important parameters relevant to the key phenomena within the ranges to investigate influences of the parameters onto the cladding surface temperature. It was confirmed that both constant C of Wallis CCFL correlation at the core exit and gas-liquid inter-phase drag in the core, as parameters that need to consider for the evaluation of safety margin, are more sensitive to the cladding surface temperature than other chosen parameters. (authors)

  16. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 3. Analysis of the OECD TMI-1 Main-Steam- Line-Break Benchmark Accident Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional (3-D) neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package is being qualified and will be used at British Energy (BE) and Tractebel Energy Engineering (TEE), independently, to analyze pressurized water reactor (PWR) transients where strong core-system interactions occur. The Organization for Economic Cooperation and Development/Nuclear Energy Agency PWR Main-Steam-Line-Break (MSLB) Benchmark problem was performed to demonstrate the capability of the coupled code package to simulate such transients, and this paper reports the BE and TEE contributions. In the first exercise, a point-kinetics (PK) calculation is performed using the RELAP5 code. Two solutions have been derived for the PK case. The first corresponds to scenario, 1 where calculations are carried out using the original (BE) rod worth and where no significant return to power (RTP) occurs. The second corresponds to scenario 2 with arbitrarily reduced rod worth in order to obtain RTP (and was not part of the 'official' results). The results, as illustrated in Fig. 1, show that the thermalhydraulic system response and rod worth are essential in determining the core response. The second exercise consists of a 3-D neutron kinetics transient calculation driven by best-estimate time-dependent core inlet conditions on a 18 T and H zones basis derived from TRAC-PF1/MOD2 (PSU), again analyzing two scenarios of different rod worths. Two sets of PANTHER solutions were submitted for exercise 2. The first solution uses a spatial discretization of one node per assembly and 24 core axial layers for both flux and T and H mesh. The second is characterized by spatial refinement (2 x 2 nodes per assembly, 48 core layers for flux, and T and H calculation), time refinement (half-size time steps), and an increased radial discretization for solution

  17. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  18. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

  19. Effect on code predictions by changing the code version of Relap5 on SBLOCA for Test 9.1B in BETHSY test facility

    International Nuclear Information System (INIS)

    This paper deals with the use of the new version of Relap-5 code on Sbloca (small break loss of coolant accident) for test 9.1 in the Bethsy integral test facility. In this analysis an updated version of the best estimate code Relap5/mod.3.3 has been used. In this code options are available for critical flow at the junction, modified Henry Fauske model (with only one discharge coefficient and thermal non equilibrium constant), and original model (with sub-cooled, two-phase and superheated discharge coefficient). Post test analyses have been carried out. This analysis has been carried out with all the procedure lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE). In order to achieve a qualified test facility nodalization both 'steady state level' and 'on transient level' qualifications are demonstrated. It is concluded that overall qualitative and quantitative accuracy of code prediction (mod.3.3) are acceptable as per UMAE. However noticeable effect of change of version of code has been observed from mod.2 to mod.3.3. Discharge coefficient for the modified Henry Fauske model is very sensitive for the break modeling point of view. The thermal non equilibrium constant for break discharge modeling is not having much effect on the analysis results. Especially when the core is refilled, code under-predicts the break flow and integral break flow consequently code over-predicts primary mass inventory. It is found that with the same input deck all the significant events and phenomena for the mod.3.3 occurring about 600 s earlier with compare to mod.2 calculation. Up to refilling of core, time sequence of the entire significant events is after the experiments in mod.2 calculation whereas in mod.3.3 it is before the experimental value. With this new version of code a better prediction for clad temperature during dry out has been observed. In both version of code prediction of results are poor for the last 2500 s of transients this may be due to large

  20. Validation of Atucha-2 PHWR helios and Relap5-3D model by Monte Carlo cell and core calculations - 335

    International Nuclear Information System (INIS)

    Within the framework of the Second Agreement 'Nucleoelectrica Argentina-SA - University of Pisa', a complex three dimensional (3D) neutron kinetics (NK) coupled thermal-hydraulic (TH) RELAP5-3D model of the Atucha 2 PHWR has been developed and validated. Homogenized cross section database was produced by the lattice physics code HELIOS. In order to increase the level of confidence on the results of such sophisticated models, an independent Monte Carlo code model, based on the MONTEBURNS package (MCNP5 + ORIGEN), has been set up. The scope of this activity is to obtain a systematic check of the deterministic codes results. This necessity is particularly felt in the case of Atucha-2 reactor modeling, since its own peculiarities (e.g., oblique Control Rods, Positive Void Coefficient) and since, if approved by the Argentinean Safety Authority, the RELAP53D 3D NK TH model will constitute the first application of a neutronic thermal-hydraulics coupled code techniques to a reactor licensing project. (authors)