J. Vedovi

University of Pisa

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Kostadin Ivanov

North Carolina State University

Francesco Saverio D'Auria

University of Pisa

A Bousbia Salah

Federal Agency for Nuclear Control

G. M. Galassi

University of Pisa

D. Panayotov

Fusion for Energy

Jiří Macek

Eastbourne District General Hospital

U. Rindelhardt

Helmholtz-Zentrum Dresden-Rossendorf

M. Yitbarek

N.S. Garis

Swedish Radiation Safety Authority

F. Reventós

Universitat Politècnica de Catalunya

Cooperative Institutions

University of Pisa

Helmholtz-Zentrum Dresden-Rossendorf

Universitat Politècnica de Catalunya

Istituto Nazionale di Fisica Nucleare, Sezione di Pisa

Institute for Nuclear Research and Nuclear Energy

Nuclear Research Institute Rez (Czechia)

Chalmers University of Technology

University of Illinois Urbana-Champaign

Urbana University

Plasma (Russia)

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Analysis of the Peach Bottom Turbine Trip 2 Experiment by Coupled RELAP5-PARCS Three-Dimensional Codes

Nuclear Science and Engineering (2004)

A Bousbia Salah J. Vedovi Francesco Saverio D'Auria Kostadin Ivanov G. M. Galassi

Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.

Neutron Transport

Thermal hydraulics

Boiling water reactor

Transient (computer programming)

Hydraulics

10.13182/nse04-a2462

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Citations (25)

Analysis of BWR OPRM plant data and detection algorithms with DSSPP

Jialun Yang J. Vedovi Aeree Chung J. F. Zino

All U.S. BWRs are required to have licensed stability solutions that satisfy General Design Criteria (GDC) 10 and 12 of 10 CFR 50 Appendix A. Implemented solutions are either detect and suppress or preventive in nature. Detection and suppression of power oscillations is accomplished by specialized hardware and software such as the Oscillation Power Range Monitor (OPRM) utilized in Option III and Detect and Suppress Solution - Confirmation Density (DSS-CD) stability Long-Term Solutions (LTSs). The detection algorithms are designed to recognize a Thermal-Hydraulic Instability (THI) event and initiate control rod insertion before the power oscillations increase much higher above the noise level that may threaten the fuel integrity. Option III is the most widely used long-term stability solution in the US and has more than 200 reactor years of operational history. DSS-CD represents an evolutionary step from the stability LTS Option III and its licensed domain envelopes the Maximum Extended Load Line Limit Analysis Plus (MELLLA +) domain. In order to enhance the capability to investigate the sensitivity of key parameters of stability detection algorithms, GEH has developed a new engineering analysis code, namely DSSPP (Detect and Suppress Solution Post Processor), which is introduced in this paper. The DSSPP analysismore » tool represents a major advancement in the method for diagnosing the design of stability detection algorithms that enables designers to perform parametric studies of the key parameters relevant for THI events and to fine tune these system parameters such that a potential spurious scram might be avoided. Demonstrations of DSSPPs application are also presented in this paper utilizing actual plant THI data. A BWR/6 plant had a plant transient that included unplanned recirculation pump transfer from fast to slow speed resulting in about 100% to {approx}40% rated power decrease and about 99% to {approx}30% rated core flow decrease. As the feedwater temperature is reduced to equilibrium conditions, the power increased from about {approx}40% to about {approx}60% with little change inflow. A THI event developed and subsequently, an OPRM initiated scram occurred with Option III. (authors)« less

Scram

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Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification

F. D’Auria A Bousbia Salah G. M. Galassi J. Vedovi F. Reventós

Neutron Transport

Transient (computer programming)

Thermal hydraulics

Hydraulics

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Citations (5)

Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP2: State-of-the-Art Report

Francesco Saverio D'Auria A Bousbia Salah J. Vedovi F. Reventós A. Cuadra

Neutron Transport

Thermal hydraulics

Hydraulics

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Numerical Modeling of Flow and Heat Transfer During Quenching for a Postulated Loss-of-Coolant Accident

Luis Ibarra J.G.M. Andersen J.T. Rambo J. Vedovi A. Petruzzi

Loss-of-coolant accident

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Analysis of the OECD/DOE/CEA VVER1000 CT-1 benchmark using Relap5/Parcs coupled codes

J. Vedovi A Bousbia Salah Francesco Saverio D'Auria G. M. Galassi Kostadin Ivanov

Benchmark (surveying)

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Citations (1)

Analyses and sensitivity studies on VVER-1000 using the best-estimate model Relap5/Parcs

J. Vedovi Kostadin Ivanov Francesco Saverio D'Auria

VVER

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V1000CT-1 Benchmark: Exercise 3 Results using coupled Three-Dimensional Kinetics/Thermal Hydraulics

Francesco Saverio D'Auria J. Vedovi Kostadin Ivanov

Benchmark (surveying)

Hydraulics

Thermal hydraulics

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Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP3: Achievements and Recommendations Report

F. D’Auria A Bousbia Salah G. M. Galassi J. Vedovi F. Reventós

Neutron Transport

Thermal hydraulics

Hydraulics

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Citations (2)

Analysis of Pressure Wave Phenomena in Steam Line Piping

10th International Conference on Nuclear Engineering, Volume 4 (2004)

J. Vedovi Michael J. Hibbard Donald Todd Kostadin Ivanov

One challenging aspect of boiling water reactor (BWR) analysis is the ability to predict the system response following the rapid closure of a steam line control or isolation valve. Of particular interest is to accurately model the effect of a pressurization wave as it transits through the reactor core. This paper describes sensitivity studies, which were performed to demonstrate the predictive capabilities of S-RELAP5 for analysis of pressure wave phenomena, and it describes the steam line models developed in support of this effort. S-RELAP5 is a RELAP5-based thermal-hydraulic system code used for realistic analyses of large break loss-of-coolant accidents (LBLOCA) in pressurized water reactors (PWRs). The code is also suitable for analyzing PWR small break LOCA and non-LOCA transients. On the extent of the analyses documented in this paper, there are not significant differences between S-RELAP5 and RELAP5. Framatome ANP is developing code models which will extend the capability of S-RELAP5 to analyze BWR transients. Within the framework of this development work, a task was established to investigate the capability of S-RELAP5 to model transients involving steam line pressure wave phenomena. An additional goal of this task was development of a steam line nodalization guideline for modeling pressurization transients. To achieve these goals, various steam line models were investigated and a series of sensitivity studies were performed using the Peach Bottom Unit 2 Turbine Trip test series as an experimental benchmark. A Turbine Trip is an anticipated operational occurrence (AOO), which, for analysis purposes, is initiated by rapid closure of the Turbine Stop Valves (TSV). The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes to the core void distribution and coolant flow. The magnitude of the resulting neutron flux transient is significantly affected by the void collapse. The performed sensitivity studies focused on steam line geometry, time step size, steam line node length, Turbine Stop Valve (TSV) model, steam Bypass Valve (BPV), and transient boundary conditions. Three models were developed for the Peach Bottom Unit 2 steam line; a single ideal steam line with no bends or elevation changes, a four steam line model of the PB2 main steam and bypass line piping, and an equivalent single steam line model of the PB2 main steam and bypass line piping. The pressure response calculated by S-RELAP5 was compared to theoretical predictions based on fundamental water-hammer equations and to experimental data from the PB2 turbine trip tests (TT1, TT2, and TT3). The results demonstrate the capability of S-RELAP5 to accurately predict pressure wave phenomena. Additional results of this work include recommendations of appropriate values for time step size and steam line node size. Models for the TSV and BPV were developed based on TT2 information, and validated through analysis of TT1 and TT3. Finally, the results of this work demonstrate the strong interaction between the steam bypass line and the main steam line, and the corresponding impact on system pressure response.

Cabin pressurization

Pressurized water reactor

Thermal hydraulics

Closure (psychology)

Boiling water reactor

10.1115/icone12-49414

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