TRETA and TIZONA fast running thermal-hydraulic codes

Annals of Nuclear Energy

Volumn 34, Issue 7, 2007, Pages 533-549

  Izquierdo, J M ^a   Sanchez M ^a   Hortal, F J ^a   Melendez E ^a   Herrero, R ^a   Queral, C ^a   Exposito A ^a   Gonzalez I ^a  

^a Presidenta de la Comisión Nacional de Radiofísica Vocal del Comité Asesor del Consejo de Seguridad Nuclear Expresidenta de la Sociedad Española de Protección Radiológica   (Spain)

Author keywords

[No Author keywords available]

Indexed keywords

BLOCK CODES; COMPUTATIONAL COMPLEXITY; COMPUTER SIMULATION; PROBABILISTIC LOGICS; PROBLEM SOLVING;

BLOCK DIAGRAMS; PROBABILISTIC SAFETY ANALYSIS (PSA); THERMAL-HYDRAULIC SECTIONS;

NUCLEAR POWER PLANTS;

EID: 34547730873     PISSN: 03064549     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.anucene.2007.02.019     Document Type: Article

References (37)

1. ANSI/ANS, 1979. Decay heat power in light power reactors. Tech. Rep. ANSI/ANS-5.1, American Nuclear Society.
2. Bestion D. The physical closure laws in the CATHARE code. Nuclear Engineering and Design 124 (1990) 229-245
3. Cojazzi, G., Meléndez, E., Izquierdo, J.M., Sánchez-Perea, M., 1992. The reliability and safety assessment of protection systems by the use of dynamic event trees. The DYLAM-TRETA Package. In: Proc. XVIII Annual Meeting Spanish Nuclear Society, Puerto de Santa María, Spain, 28-30 October 1992.
4. Devooght J., Izquierdo J.M., and Meléndez E. Relationships between probabilistic dynamics and event trees. Reliability Engineering and System Safety 52 (1996) 197-209
5. EPRI, 1991. RETRAN-03, a program for transient thermal-hydraulic analysis of complex fluid flow systems. Tech. Rep. NP-7450-CCML, American Nuclear Society.
6. Expósito, A., 2007. Development of a simulation methodology for accident sequences in light water nuclear power plants considering operator actions. Ph.D. thesis, ETSI Minas (UPM) (in spanish).
7. Expósito, A., Queral, C., 2005. Modifications made to the TIZONA code module PIPEIB. Tech. Rep. DSE-08/2005 (in Spanish).
8. Expósito, A., Queral, C., 2006. PWR-W plant model for TRETA code. Ver. 3. Tech. Rep. (in Spanish).
9. Expósito, A., Queral, C., Quiroga, J.A., Ibarra, A., Hortal, J., 2004. Simulation of accident sequences including emergency operating procedures. In: Proceedings of International Congress on Advanced Nuclear Power Plants. Pittsburgh, PA USA, June 13-17.
10. Expósito, A., Hortal, J., Queral, C., Quiroga, J.A., Hulsun, J.E., 2005. A new approach to incorporate operator actions in the simulation of accident sequences. European Nuclear Conference, Versalles, France, 11-14 December 2005.
11. Expósito, A., Queral, C., Carballo, J., 2005. Cofrentes (BWR/6) plant model for TIZONA code. Tech. Rep. DSE-11/2005 (in Spanish).
12. Expósito, A., Hortal, J., Queral, C., Quiroga, J.A., Ibarra, A., Hulsun, J.E., Meléndez, E., Elorza, F.J. Analysis and verification of emergency operating procedures through the integrated simulation of plant and operator actions. In: IAEA Topical Meeting on Advanced Safety Assessment Methods for Nuclear Reactors. Korea Institute of Nuclear Safety. Daejon, Republic of Korea, 30 October-2 November 2007, submitted for publication.
13. Giles, M.M. et al., 1992. TRAC-BF1/MOD1: An advanced best-estimate computer program for bwr accident analysis. Tech. Rep. NUREG/CR-4356, American Nuclear Society.
14. Herrero R., 1999. A standardised methodology for the linkage of computer codes. Application to RELAP5/Mod3.2, NUREG/IA-0179.
15. Herrero, R., Izquierdo, J.M., 2004. Development of a computer tool for in-depth analysis and postprocessing of the RELAP 5 thermalhydraulic code. Tech. Rep. 31.2004, Consejo de Seguridad Nuclear.
16. Hortal, J. 1996. Simulation of operating procedures as a tool for NPP procedure verification, Enlarged HRP meeting, Loen (Norway) 19-24th May, 1996.
17. Hortal J., and Izquierdo J.M. Application of the integrated safety assessment methodology to the protection of electric systems. Reliability Engineering and System Safety 52 (1996) 315-326
18. Izquierdo, J.M., Sánchez-Perea, M., Hortal, J., Meléndez, E., 1996. New Techniques for the Application of Dynamic Event Trees to the Interaction between Electric Grids and Generating Stations. In: ESREL'96-PSAM III, Crete, Greece, June 1996.
19. Izquierdo, J.M. et al., 2002. An integrated PSA approach to independent regulatory evaluations of nuclear safety assessments of spanish nuclear power plants. Tech. Rep. 28.2002, Consejo de Seguridad Nuclear.
20. Izquierdo, J.M., Sánchez-Perea, M., 1990. DYLAM-TRETA: An approach to protection systems software analysis. In: Advanced Systems Reliability Modeling, Proc. Ispra course held at ETSI Navales, Madrid, Spain, September 1988.
21. Izquierdo J.M., and Sánchez-Perea M. Application of the integrated safety assessment methodology to the emergency procedures of a SGTR of a PWR. Reliability Engineering and System Safety 45 (1994) 159-173
22. Izquierdo J.M., Hortal J., Veci I., Perez J., Pelayo F., and Sánchez-Perea M. TRETA: a general purpose simulation code with application on nuclear power plant transients. Journal of the Spanish Nuclear Society 58 (1987) 41-47 (in Spanish)
23. Izquierdo, J.M., Sánchez-Perea, M., Cacciabue, P.C. 1988. Dynamic reliability as a tool for the assessment of protection systems software analysis, presented at the NUCSAFE ENS/ANS Conference, Avignon, France, October 1988.
24. Izquierdo, J.M., Hortal, J., Meléndez, E., Sánchez-Perea, M., 1996. Automatic generation of dynamic event trees: a tool for integrated safety assessment (ISA). In: Aldemir, T. et al. (Ed.), Reliability and Safety Assessment of Dynamic Process Systems, NATO ASI series F, vol. 120. Springer Verlag, Berlin, 1994, Reliability Engineering and System Safety, 52, 1996.
25. Izquierdo, J.M., Queral, C., Herrero, R., Hortal, J., Sánchez-Perea, M., Meléndez, E., Muñoz, R., 2000. Role of Fast Running TH Codes and Their Coupling with PSA Tools. NEA/CSNI/R(2001)2, vol. 2, Workshop Proceedings, Barcelona (Spain) 10-13 April 2000.
26. Lellouche, G., Chexal, B.A., 1986. Full-range drift flux correlation for vertical flows, rev. 1. Tech. Rep. NP-3969-SR, EPRI.
27. Meléndez, E, Izquerdo, JM, Sánchez-Perea, M, Hortal, J, Pérez-Mulas, A., 1999. Tree Simulation techniques for Integrated Safety Assessment, Conferencia CSNI Specialist Meeting on Simulators and Plant Analysers, Espoo, Finland.
28. Muñoz, R., Mínguez, E., Meléndez, E., Izquierdo, J.M., Sánchez-Perea, M., 1999. DENDROS: A Second Generation Scheduler for Dynamic Event Trees, M&C'99, Madrid.
29. Queral, C., 1995. Algorithm for the solution of the two-phase flow conservation equations. application to nuclear power plants. Ph.D. thesis, Departamento de Física Fundamental. UNED.
30. Queral, C., García, M.A., 1995. Transient Simulation Practices in nuclear power plants. Nuclear tests. Departament of Energy Systems. ETSI de Minas (in Spanish).
31. Sánchez-Perea, M., 1996. Analysis and simulation of networks of linear subsystems by means of series expansions of laguérre functions. Ph.D. thesis, ETSI Industriales de ICAI (in Spanish).
32. Sánchez-Perea, M., Melara J., 1996. Extending PSA to accident management. The case of the steam generator tube rupture (SGTR) emergency operating procedures assessment. In: International Conference on Nuclear Engineering (ICONE-IV) ASME meeting. New Orleans, March 10-14 1996.
33. The RELAP5 Code Development Team. June 1995. RELAP5/MOD3 code manual, volume I: Code structure, system models and solution methods. Tech. Rep. NUREG/CR-5535,INEL-95/0174. Idaho National Engineering Laboratory.
34. Wulff, W., 1998. Thermal hydraulics od two-phase flow in complex systems. In: ANS (Ed.), 1998 ANS Winter Meeting.
35. Wulff, W., Cheng, H.S., Lekach, S., Mallen, A., August 1984. The BWR plant analyzer. Tech. Rep. NUREG/CR-3943. BNL-NUREG-51812, Brookhaven National Laboratory.
36. Wulff, W. et al., 1984. A description and assessment of RAMONA-3b mod.0 cycle4: A computer code with three-dimensional neutron kinetics for BWR system transients. Tech. Rep. NUREG/CR-3664, US NRC.
37. Zuber N., and Findlay J. Average volumetric concentration in two-phase flow systems. Journal of Heat Transfer. Transactions of the ASME (1965) 453-468