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Wednesday, November 11, 2020 | History

4 edition of LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3 found in the catalog.

LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3

LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3

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Published by Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission in Washington, DC .
Written in English

    Subjects:
  • Pressurized water reactors -- Loss of coolant -- Computer simulation

  • Edition Notes

    Statementprepared by J.I. Sánchez, C.A. Lage, T. Núñuz
    SeriesInternational agreement report -- NUREG/IA-0195
    ContributionsLage, C. A, Núñez, T, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research, Empresa Nacional del Uranio (Spain)
    The Physical Object
    FormatMicroform
    Paginationix, 61 p.
    Number of Pages61
    ID Numbers
    Open LibraryOL13628519M
    OCLC/WorldCa53244071


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LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3 Download PDF EPUB FB2

This report documents the analysis of a postulated Large Break Loss-of-Coolant Accident (LBLOCA) in a Westinghouse 3-Loop PWR design analysed using the "Best Estimate" code RELAP5/MOD3.

This LBLOCA calculation represents ENUSA's contribution to the "Code Assessment and Maintenance Program" (CAMP). LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3 Prepared by J.I. Sinchez, C.A.

Lage, T. Nfifiez Empresa Nacional del Uranio S.A. Santiago Rusinol, 12 Madrid SPAIN Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC January Prepared as part of Get this from a library.

LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3. [J I Sánchez; C A Lage; T Núñez; U.S. Nuclear Regulatory Commission. Office   Title: NUREG/IA, "International Agreement Report, LBLOCA analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3." Created Date: 9/12/ PM   [10] ENUSA, "LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3", NUREG-IA, [11] Westinghouse, "Development and Qualification of a GOTHIC Containment Evaliation Model for the Kewaunee Nuclear Power Plant," [12] F.

Rahn, "GOTHIC Tech Manual," LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3. Nuclear Regulatory Commission, U.S.

This book emphasizes the prevention   Results of the LBLOCA analysis have been reported in Chapter 15 of DCD on Accident Analysis (Westinghouse, a) using WCOBRA–TRAC code. The reported value of PCT from this analysis is K with core divided into 10 volumes axially. LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3 book et al.

() also analyzed an LBLOCA for AP using TRACE V patch 2 code. TRACE model used a 17   LBLOCA: Fases En los 2 ó 3 primeros minutos después de un LOCA hay cuatro fases: 1. Blowdown: el refrigerante sale del SRR. Bypass: cuando el vapor ascendiente del downcomer impide inyectar el agua de refrigeración en la parte inferior de la vasija 3.

Refill: cuando el agua del sistema de refrigeración de Analysis of Inadvertent Pressurizer Spray Valve Opening Real Transient with RELAP5/MOD NUREG/IA LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3: NUREG/IA Analysis of PANDA Experiments P3 and P6 Using RELAP5/MOD NUREG/IA A model of VVER reactor based on Unit 1 of BNPP has been developed for the RELAP5/MOD thermal-hydraulics code consists of 4-loop primary and secondary systems with all their relevant sub   LBLOCA is analyzed in two beyond design basis accident scenarios using RELAP5/MOD • A RELAP5 model of VVER reactor based on Unit 1 of BNPP is developed.

• The BDBA scenarios are LBLOCA with SBO and LBLOCA with pump re-circulation blockage. • Effect of water inventory of ECCS passive parts on core cooling is investigated. • Analysis in a W estinghouse PWR 3-loop design, RELAP5/MOD3'.

NUREG The main objective of this project is to obtain a standard PWR Westinghouse as well as an AP containment model for a Voltage based repair criteria for Westinghouse steam generator tubes affected by outside diameter stress corrosion cracking.

Published: () Steam line break analysis using RELAP5/MOD for steam generator blowdown load assessment / by: Jung, Han-Sik, Published: () Westinghouse, Westinghouse AP design control document Rev. Chapter Accident Analysis, MLA, Integrated Uncertainty Analysis using RELAP/SCDAPSIM/MOD 27   Design and define scenarios: The selected plant design is a PWR-W with 3 loops and the sequence will be a LONF-ATWS.

However, as already mentioned, this analysis is suitable for a 4-Loop plant, as the phenomena are similar for most sequences, (Kemper et al., ). @article{osti_, title = {Realistic Small- and Intermediate-Break Loss-of-Coolant Accident Analysis Using WCOBRA/TRAC}, author = {Bajorek, Stephen M and Petkov, Nikolay and Ohkawa, Katsuhiro and Kemper, Robert M and Ginsberg, Arthur P}, abstractNote = {Since the Appendix K Rulemaking change, there has been significant interest in the development of codes and LBLOCA Analysis in a Westinghouse PWR 3-Loop Design Using RELAP5/MOD3: Jan ML NUREG/CR R1: Manual of Respiratory Protection Against Airborne Radioactive Material: Jan ML NUREG/CR V1 R2: SCDAP/RELAP5/MOD Code Manual: Code Architecture and Interface of Thermal Hydraulic and Core Behavior Models: Jan The aim of this analysis has been to evaluate the risk contribution due to the offsite dose and the core damage in case of Steam Generator Tube Rupture (SGTR) sequences at full power in a 3-loop Best Estimate Analysis of Maanshan PWR LBLOCA by TRACE Coupling With DAKOTA in both the experimental programs and the RELAP5/MOD3 systems analysis computer code.

a Westinghouse   Zion Unit 1 was a four loop Westinghouse design PWR, located in Illinois, USA. The plant rated thermal power was MW. It ceased operation in It is now permanently shut down. The steady-state conditions are listed in Table 2, extracted from   A LBLOCA (large-break loss of coolant accident) similarity with the APR was assessed for a reduced-height integral effect test loop design for PWRs (pressurized water reactors) during the basic design of the ATLAS facility (Park et al., a), and the simulation capability of the ATLAS for major design-basis accidents, including a large Aplicación de difusión de la investigación en la UPM.

S2i Observatorio de investigación @ UPM con la colaboración del Consejo Social ?pageac=&idInvestigador= Conference: Analysis of the RCS over cooling issue of a Westinghouse four-loop vintage plant using RELAP5/MOD3 @article{osti_, title = {Quantifying reactor safety margins: Application of CSAU (Code Scalability, Applicability and Uncertainty) methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes}, author = {Wulff, W and Boyack, B E and Duffey, R B and Griffith, P and Katsma, K R and Lellouche, G S and Levy, S and Rohatgi, U S and Wilson RELAP5 analysis of mitigation strategy for extended blackout power condition in PWR / E-ZBorrow is the easiest and fastest way to get the book you want (ebooks unavailable).

Use ILLiad for articles and chapter scans. 2 run 2 with RELAP5 and TRACE codes application to a PWR NPP model / by: LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3 / by: Sánchez, J.

I., et al. Published: () Hydraulic transport of coating debris a subtask of GSI / Published: () Assessment of RELAP5/MOD3 with the SNUF test simulating hot leg break LOCA in the view of mass and energy release analysis / by: Hong, S.

Joon, Published: () Simulation of LOCA 6" and LOCA 2" transients in the RHR of a PWR under low power conditions using RELAP5/MOD / by: Martorell, S., Published: ()   The U.S. Nuclear Regulatory Commission (NRC) revised the emergency core cooling system (ECCS) licensing rule to allow the use of best-estimate (BE) computer codes, provided that the uncertainties of the calculations are quantified and used in the   for the LBLOCA in an OPR and the other is for a LBLOCA in a Westinghouse four-loop PWR.

From the review, the 31 candidate parameters were set up and the probability distributions, uncertainty ranges and reference values for all parameters ?. @article{osti_, title = {Achieving 95% probability level using best estimate codes and the code scaling, applicability and uncertainty (CSAU) (Code Scaling, Applicability and Uncertainty) methodology}, author = {Wilson, G E and Boyack, B E and Duffey, R B and Griffith, P and Katsma, K R and Lellouche, G S and Rohatgi, U S and Wulff, W and Zuber, N}, abstractNote = {Issue of a revised The boundary test conditions were defined based on PWR LBLOCA analysis by RELAP5/MOD code considering typical Japanese safety analysis conditions.

Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study.

The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the exercise. The BEMUSE (Best Estimate Methods plus Uncertainty and The third step is the main aim of this paper and consists of a continuation of the previous projects in the field of NPP analysis.

The aim of this paper is to study SBLOCA transients with boron dilution in PWR. The chosen NPP was Ascó-2 which is a 3-loop,6 MWth Westinghouse :// The task of regulatory body staff reviewing and assessing a realistic large break loss-of-coolant accident evaluation model is discussed, facing the actual regulatory licensing environment related to the acceptance of the analysis of emergency core cooling system performance.

Especially, focus is directed to the question of how to fulfill the requirement of quantifying the uncertainty in the Fouet, Fabrice, and Probst, Pierre. "Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V Code." Proceedings of the 21st International Conference on Nuclear Engineering.

Volume 4: Thermal Hydraulics. Chengdu, China. July 29–August 2, VT09A :// Assessment of RELAP5/MOD Using LOFT Large Break International Agreement Report NUREG/IA International Agreement Report Assessment of RELAP5/MOD Using LOFT Large Break LOCA Test, LP- 6 Prepared by T S. Choi, J.

Lee, B. Park, C. Cho, J.Y. Park/KNFC Y. Bang, S. W Seul, H. Kim/KINS Korea Nuclear Fuel Company DogJin-Dong Yusong Gu, Daejeon City 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 :// AP is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation.

Its core design is very similar to 3-loop PWR with fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using :// A method for the assessment of portable pump flow rates for a steam generator (SG) makeup is proposed.

The RELAP5/MOD computer code and input model of a two-loop pressurized water reactor (PWR) is used for analyses, assuming different injection start times, flow rates, and reactor coolant system (RCS) :// For calculations the RELAP5/MOD 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. The passive safety systems of AP are designed to operate automatically at desired set-points. However, the unavailability or failure to operate of any of the passive safety systems will change the accident sequence and may affect reactor safety.

The analysis in this study is based on some hypothetical scenarios, in which the passive safety system failure is considered during the loss of The loop has to keep its full integrity after decontamination, so the chemicals can’t be too aggressive.

The second part comprises an analysis of different technical designs for the decontamination installation. Important is the provision of a thorough ://  Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study.

The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the