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Multiphase Science and Technology

年間 4 号発行

ISSN 印刷: 0276-1459

ISSN オンライン: 1943-6181

SJR: 0.144 SNIP: 0.256 CiteScore™:: 1.1 H-Index: 24

Indexed in

APPROACHES ADOPTED FOR CRITICAL HEAT FLUX EVALUATION DURING TRANSIENT USING SYSTEM ANALYSIS CODE RELAP-5 FOR KKNPP

巻 31, 発行 4, 2019, pp. 359-370
DOI: 10.1615/MultScienTechn.2020031213
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要約

Critical heat flux (CHF) condition is an important factor in heat-flux-controlled systems such as nuclear reactors since the temperature increase can threaten the physical integrity of the heated surface. KKNPP has two operating VVER-1000 reactors. VVER-1000 is a pressurized water reactor (PWR) incorporating advanced safety features including passive ones. In a PWR, CHF signifies an important thermal-hydraulic safety limit, which has to be maintained within prescribed limits for all operating and transient conditions. Various postulated initiating events have been analyzed using thermal hydraulic computer code RELAP-5/MOD 3.2. The calculations are performed from the viewpoint of checking the departure from nucleate boiling ratio (DNBR) during the whole transient. In the analysis, the DNBR is calculated in two ways: RELAP look-up tables and externally coupled designer-specified correlations such as Gidropress, Smolin, and Kutateladze. The code provides DNBR value at different nodes, and the minimum value among these nodes is used in analysis. Using a conservative approach, minimum DNBR has been established for anticipated operational occurrences under various functional categories, including (1) increase in heat removal through secondary circuit, (2) decrease in heat removal through secondary circuit, and (3) loss of primary coolant flow. The DNBR values for each type of event are studied and checked with acceptance criteria (DNBR > 1.0).

参考
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  8. Ransom, V.H., Trapp, J.A., and Wagner, R.J., RELAP-5/MOD 2 Code Manual, Volumes I and II, Idaho Falls, ID, EG&G Idaho, Inc., Rep. NUREG/CR-4312, EGG-2396, August and December 1985; revised April 1987.

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によって引用された
  1. Mehta Manish, Chaudhary Sanuj, Biswangri Anirban, Krishna Kumar P., Pandey Y. K., Biswas Gautam, Safety Analysis of Loss of NPP Off-Site Power with Failure of Reactor SCRAM (ATWS) for VVER-1000, in Proceedings of the 7th International Conference on Advances in Energy Research, 2021. Crossref

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