Volume 20, Issue 3 (March 2020)                   Modares Mechanical Engineering 2020, 20(3): 565-573 | Back to browse issues page

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Karamimehr Z, Rabiee A. Operator Action analysis in Main Steam Line Break Connected to Turbine and Loss of Steam Generator Feedwater in Bushehr powerplant. Modares Mechanical Engineering 2020; 20 (3) :565-573
URL: http://mme.modares.ac.ir/article-15-30402-en.html
1- Mechanical Engineering Faculty, University of Shiraz, Shiraz, Iran , karamimehrzahra@gmail.com
2- Mechanical Engineering Faculty, University of Shiraz, Shiraz, Iran
Abstract:   (4803 Views)
The operation of power generation cycles and their related events are one of the main issues in the field of safety of power plants. If these events are not properly managed for any reason, the consequences will be irreparable. In the meantime, the operator action can be one of the most effective factors in the management of the accident. In this research, the operator action has been evaluated in the main steam line break connected to the turbine and total loss of steam generator feed water for the Bushehr power plant. Firstly, the data has been validated in both steady and transient states with the final safety analysis report of the power plant of Bushehr as a reliable reference. The results indicate a good agreement with the final safety analysis report. In the next step, the operator action has been evaluated to mitigate the thermohydraulic parameters, including temperature and pressure. Finally, by performing an operator sensitivity analysis in the main steam line break connected to the turbine followed by total loss of steam generator feed water, the maximum possible time for operator intervention has been estimated 76 minutes.
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Article Type: Original Research | Subject: Thermal Power Plant
Received: 2019/02/13 | Accepted: 2019/05/14 | Published: 2020/03/1

References
1. Atomic Energy Organization of Iran. Final Safety Analysis Report (FSAR) for BNPP: Accident Analysis [Report]. Unknown Publisher; 2007. [Link]
2. nrc.gov [Internet]. Washington: United States Nuclear Regulatory Commission; 2016 [Unknown cited] Available from: Not Found [Link]
3. Svenson O. A decision theoretic approach to an accident sequence: When feedwater and auxiliary feedwater fail in a nuclear power plant. Reliability Engineering & System Safety. 1998;59(2):243-252. [Link] [DOI:10.1016/S0951-8320(97)00127-0]
4. Gencheva R, Stefanova A, Groudev P. Investigation of steam line break accident during the development of emergency operating procedures for VVER440/V230. Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee: 2002. [Link]
5. Muellner N, Giannotti W, D'Auria F. Investigation of a possible emergency procedure for the VVER1000 NPP in case of total loss of feed water and a main steam line break. Proceedings of the 12th International Conference on Nuclear Engineering; Arlington, Virginia: 2004, April 25-29. New York; American Society of Mechanical Engineers; 2004. [Link] [DOI:10.1115/ICONE12-49036]
6. Pavlova MP, Groudev PP, Stefanova AE, Gencheva RV. RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at unit 6 of Kozloduy NPP. Nuclear engineering and design. 2006;236(3):322-331. [Link] [DOI:10.1016/j.nucengdes.2005.08.009]
7. Bucalossi A, Del Nevo A, Moretti F, D'Auria F, Elkin IV, Melikhov OI. Investigation of accident management procedures related to loss of feed water and station blackout in PSB-VVER integral test facility. Nuclear Engineering and Design. 2012;250:633-645. [Link] [DOI:10.1016/j.nucengdes.2012.06.027]
8. Lim J, Choi SW, Yang J, Lee DY, Rassame S, Hibiki T, Ishii M. Assessment of passive safety system performance under main steam line break accident. Annals of Nuclear Energy. 2014;64:287-294. [Link] [DOI:10.1016/j.anucene.2013.05.032]
9. Pavlova M, Andreeva M, Groudev P. Steam line break investigation at full power reactor for VVER-1000/V320. Nuclear Engineering and Design. 2015;285:65-74. [Link] [DOI:10.1016/j.nucengdes.2015.01.006]
10. Atomic Energy Organization of Iran. Final safety analysis report (FSAR) for BNPP: Introduction and general description of NPP [Report]. Unknown Publisher; 2007. [Link]
11. US Nuclear Regulatory Commission. RELAP5/MOD3.2 Code Manual: Code structure, system models, and solution methods. Idaho: National Engineering Laboratory; 2001. [Link]
12. Mozaffari M. Simulation of the effect of a small break loss of coolant accident (SBLOCA) transient on VVER-1000 reactor pressure vessel parameters [Dissertation]. Shiraz: University of Shiraz; 2006. [Link]
13. US Nuclear Regulatory Commission. RELAP5/MOD3.2 Code Manual: Code structure, system models, and solution methods. Idaho: National Engineering Laboratory; 2001. [Link]
14. Allis-Chalmers Manufacturing Company. Steam System and feed water system failures [Internet]. Washington: United States Nuclear Regulatory Commission; 1962 [Unknown cited]. Available from: https://digital.library.unt.edu/ark:/67531/metadc872288/m1/1/ [Link]
15. Luitjens J. Code calibration and validation framework. Idaho: Idaho National Laboratory; 2011. [Link]
16. Kliment T, Kvizda B, Zold T. Small break LOCA analysis of Mochovce NPP VVER-440/213 with operator action. Proceeding of the 6th International Information Exchange Forum on Safety Analysis Nuclear Power Plants VVER RBMK Types (Forum-6), 2002, April 8-12, Kiev, Ukraine; 2002. [Link]

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