Abstract
Generally, the thermal hydraulic (TH) codes need the results of Neutron Kinetics (NK) codes providing the reactivity properties to calculate neutron flux. Then the TH codes perform the safety analyses obtaining the responses of pressure, temperature, or water level. Two kinds of different codes calculate different physical behaviors sequentially and separately. Simultaneously computing thermal hydraulic and neutron kinetics behaviors can enhance the accuracy of the analysis. Hence, it is crucial to develop the TH-NK coupled model. This study presents the capability of the TH-NK coupled model, developed by TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator), for the BWR-4 nuclear power plant. The establishment of the TRACE/PARCS model presented the nodal and component modeling methodologies. This model was used to simulate two startup tests of high power level system transients. Principal system responses, calculated by the TRACE/PARCS model, were compared with the measured data in startup tests and the results of the point kinetic calculation of the TRACE (TRACE/PK) to evaluate the model. The evaluation shows that the TRACE/PARCS model can simulate the interaction between thermal hydraulic and neutron kinetics phenomena and predict the transients suitably. Through the comparison, the TRACE/PARCS model can be confident doing the analyses of normal and abnormal operational transients to predict the transient responses.
Abstract
Im Allgemeinen benötigen thermohydraulische (TH) Codes als Eingabedaten die Ergebnisse von Neutronenkinetik (NK) Codes, die die Reaktivitätseigenschaften zur Berechnung des Neutronenflusses liefern. Dann führen die TH-Codes die Sicherheitsanalysen durch, um die Parameter wie Druck, Temperatur oder Wasserstand zu berechnen. Zwei verschiedene Codes berechnen nacheinander und getrennt verschiedene physikalische Verhaltensweisen. Die gleichzeitige Berechnung des thermohydraulischen und neutronenkinetischen Verhaltens kann die Genauigkeit der Analyse verbessern. Daher ist es wichtig, ein gekoppeltes TH-NK-Modell zu entwickeln. In dieser Studie wird die Leistungsfähigkeit des gekoppelten TH-NK-Modells, das von TRACE (TRAC/RELAP Ad-vanced Computational Engine) und PARCS (Purdue Advanced Reactor Core Simulator) entwickelt wurde, für das Kernkraftwerk BWR-4 vorgestellt. Bei der Erstellung des TRACE/PARCS-Modells wurden die Methoden der Knoten- und Komponentenmodellierung vorgestellt. Dieses Modell wurde zur Simulation von zwei Anfahrversuchen mit Systemtransienten auf hohem Leistungsniveau verwendet. Die mit dem TRACE/PARCS-Modell berechneten Hauptreaktionen des Systems wurden mit den Messdaten der Anfahrversuche und den Ergebnissen der punktkinetischen Berechnung von TRACE (TRACE/PK) verglichen, um das Modell zu bewerten. Die Auswertung zeigt, dass das TRACE/PARCS-Modell die Wechselwirkung zwischen thermohydraulischen und neutronenkinetischen Phänomenen simulieren und die Transienten angemessen vorhersagen kann. Der Vergleich zeigt, dass das TRACE/PARCS-Modell bei der Analyse von normalen und anormalen Betriebstransienten zur Vorhersage der transienten Reaktionen zuverlässig ist.
References
1 Feltus, M. A.: Coupled 3-D kinetics thermal-hydraulic analysis of Hot Zero Power main steam line breaks using RETRAN and STAR codes. Nuclear Engineering and Design 146 (1994) 439–450, DOI:10.1016/0029-5493(94)90349-210.1016/0029-5493(94)90349-2Search in Google Scholar
2 Ivanov, K. N.; Todorova, N. K.; Sartori, E.: Using the OECD/NRC Pressurized Water Reactor Main Steam Line Break Benchmark to Study Current Numerical and Computational Issues of Coupled Calculations, Nuclear technology 142 (2003) 95–115, DOI:10.13182/NT03-A337610.13182/NT03-A3376Search in Google Scholar
3 Ivanov, B.D.; Ivanov, K. N.; Royer, E.; Aniel, S.; Bieder, U.; Kolev, N.; Groudev, P.: OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis. Progress in Nuclear Energy 48 (2006) 728–745, DOI:10.1016/j.pnucene.2006.06.00210.1016/j.pnucene.2006.06.002Search in Google Scholar
4 Kozmenkov, Y.; Kliem, S.; Grundmann, U.; Rohde, U.; Weiss, F. P.: Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET. Nuclear Engineering and Design 237 (2007), 1938–1951, DOI:10.1016/j.nucengdes.2007.02.02110.1016/j.nucengdes.2007.02.021Search in Google Scholar
5 Kliem, S.; Danilin, S.; Hämäläinen, A.; Hádek, J.; Keresztúri, A.; Siltanen, P.: Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of main steam line break benchmarks in a NPP with VVER-440 reactor. Nuclear Science and Engineering 157 (2007) 280–298, DOI:10.13182/NSE07-A272810.13182/NSE07-A2728Search in Google Scholar
6 Vanttola, T.; Hämäläinen, A.; Kliem, S.; Kozmenkov, Y.; Weiss, F. P.; Keresztúri, A.; Danilin, S.: Validation of coupled codes using VVER plant measurements, Nuclear Engineering and Design 235 (2005) 507–519, DOI:10.1016/j.nucengdes.2004.08.04710.1016/j.nucengdes.2004.08.047Search in Google Scholar
7 Hämäläinen, A.; Kyrki-Rajamäki, R.; Mittag, S.; Kliem, S.;Weiss, F. P.; Langenbuch, S.; Hegyi, G.: Validation of coupled neutron kinetic/ thermal-hydraulic codes Part 2: Analysis of a VVER-440 transient (Loviisa-1). Annals of Nuclear Energy 29 (2002) 255–269, DOI:10.1016/S0306-4549(01)00039-110.1016/S0306-4549(01)00039-1Search in Google Scholar
8 Mittag, S.; Kliem, S.; Weiss, F. P.; Kyrki-Rajamäki, R.; Hämäläinen, H.; Langenbuch, S.; Panayotov, D.: Validation of coupled neutron kinetic/thermal-hydraulic codes Part 1: Analysis of a VVER-1000 transient (Balakovo-4), Annals of Nuclear Energy 28 (2001) 857–873, DOI:10.1016/S0306-4549(00)00095-510.1016/S0306-4549(00)00095-5Search in Google Scholar
9 Grundmann, U.; Kliem, S.; Rohde, U.: Analysis of the boiling water reactor turbine trip benchmark with the codes DYN3D and ATHLET/DYN3D. Nuclear Science and Engineering 148 (2004) 226–234, DOI:10.13182/NSE04-A245310.13182/NSE04-A2453Search in Google Scholar
10 Lee. D.; Downar, T. J.; Ulses, A.; Akdeniz, B.; Ivanov, K. N.: Analysis of the OECD/NRC BWR turbine trip transient benchmark with the coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. Nuclear Science and Engineering 148 (2004) 291–305, DOI:10.13182/NSE04-A245910.13182/NSE04-A2459Search in Google Scholar
11 Cheng, L. Y.; Baek, J. S.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.: TRACE simulation of BWR anticipated transient without scram leading to emergency depressurization. Embedded Topical Meeting on Advances in Thermal Hydraulics (ATH 2014), Reno, Nevada, June 2014Search in Google Scholar
12 Lin, H. T.; Wang, J. R.; Chen, H. C.; Shih, C.: The development and assessment of TRACE/PARCS model for Lungmen ABWR. Nuclear Engineering and Design 273 (2014) 241–250, DOI:10.1016/j.nucengdes.2014.03.02710.1016/j.nucengdes.2014.03.027Search in Google Scholar
13 Elsawi, M. A.; Bin Hraiz, A. S.: Benchmarking of the WIMS9/ PARCS/TRACE code system for neutronic calculations of the Westinghouse AP1000TM reactor. Nuclear Engineering and Design 293 (2015) 249–257, DOI:10.1016/j.nucengdes.2015.08.00810.1016/j.nucengdes.2015.08.008Search in Google Scholar
14 Mascari, F.; Vella, G.; Casamassima, V.; Parozzi, F.: Analyses of Trace-Parcs Coupling Capability. Nuclear Energy for New Europe, Bovec, Slovenia, September 12–15, 2011, p. 816.1–816.9Search in Google Scholar
15 Alzaben, Y.; Sanchez-Espinoza, V. H.; Stieglitz, R.: Analysis of a steam line break accident of a generic SMART-plant with a boron-free core using the coupled code TRACE/PARCS. Nuclear Engineering and Design 350 (2019) 33–42, DOI:10.1016/j.nucengdes.2019.05.00210.1016/j.nucengdes.2019.05.002Search in Google Scholar
16 Busquim e Silva, R. A.; Shirvan, K.; Cruz, J. J.; Marques, R. P.; Marques, A. L.F.; Piqueira, J. R. C.: Advanced method for neutronics and system code coupling RELAP, PARCS, and MATLAB for instrumentation and control assessment. Annals of Nuclear Energy 140 (2020) 107098, DOI:10.1016/j.anucene.2019.10709810.1016/j.anucene.2019.107098Search in Google Scholar
17 U.S. NRC: TRACE V5.840 User’s Manual Volume 1: Input Specification. (2014)Search in Google Scholar
18 U.S. NRC: TRACE V5.840 User’s Manual Volume 2: Modeling Guidelines. (2014)Search in Google Scholar
19 U.S. NRC: TRACE V5.0 ASSESSMENT MANUAL Main Report. (2008)Search in Google Scholar
20 Downar, T. J.; Xu, Y.; Seker, V.: PARCS v3.0U.S. NRC Core Neutronics Simulator USER MANUAL. (2013)Search in Google Scholar
21 Downar, T. J.; Xu, Y.; Seker, V.: PARCS v3.0U.S. NRC Core Neutronics Simulator Theory Manual. (2012)Search in Google Scholar
22 Taiwan Power Company: Chinshan nuclear power station units 1 and 2 final safety analysis report. (2008)Search in Google Scholar
23 AREVA NP Inc.: Chinshan unit 1 cycle 27 principal plant parameters. (2012).Search in Google Scholar
24 Cohen, T. C.; Kumar, G. V.: Chinshan unit 1 startup test results final summary report. (1979)Search in Google Scholar
25 Huang, J. J.; Chen, H. C.;Wang, J. R.; Liao, L. Y.; Lin, H. T.; Shih, C.: The load rejection transient analysis of Chinshan NPP (BWR/4) using TRACE/PARCS. The 22nd International Conference on Nuclear Engineering (ICONE 2014) p. 30858, DOI:10.1115/ICONE22-3085810.1115/ICONE22-30858Search in Google Scholar
© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
