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Multiphase Science and Technology
SJR: 0.183 SNIP: 0.483 CiteScore™: 0.5

ISSN Imprimer: 0276-1459
ISSN En ligne: 1943-6181

Multiphase Science and Technology

DOI: 10.1615/MultScienTechn.2020031213
pages 359-370

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

Manish Mehta
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094
Sanuj Chaudhary
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094
Anirban Biswangri
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094
P. Krishna Kumar
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094
Y. K. Pandey
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094
Gautam Biswas
Directorate of Engineering (LWR), Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai-400094

RÉSUMÉ

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).

RÉFÉRENCES

  1. Asmolov, V.G., Development of NPP Design based on the VVER Technology, MDEP Conf. on New Reactor Activities, Paris, France, pp. 103-115, 2009.

  2. Chaudhary, S., Krishna Kumar, P., Pandey, Y.K., and Biswas, G., Thermal-Hydraulic Analysis of Partial Loss of Forced Reactor Coolant Flow with Non-Uniform and Asymmetric Loop Flow Mixing in VVER-1000, Proc. of the 7th Int. 45th National Conf. on Fluid Mechanics and Fluid Power, Bombay, India, paper no. 71,2018.

  3. Glasstone, S. and Sesonske, A., Nuclear Reactor Engineering, New Delhi, India: CBS Publication, vol. 2, pp. 543-550,2004.

  4. Groeneveld, D.C., Cheng, S.C., and Doan, T., AECL-UO Critical Heat Flux Look-Up Table, Heat Transf. Eng., vol. 7, pp. 46-42,1986.

  5. Katto, Y., Critical Heat Flux, IntJ. Multiphase, vol. 20, pp. 53-90, 1994.

  6. Mozafari, M.A. and Faghihi, F., Design of Annular Fuels for a Typical VVER-1000 Core: Neutronic, Investigation, Pitch Optimization and MDNBR Calculation, Annals. Nuclear Energy, vol. 60, pp. 226-234, 2013.

  7. Nukiyama, S., The Maximum and Minimum Values of the Heat Q Transmitted from Metal to Boiling Water under Atmospheric Pressure, Int. J. Heat Mass Transf., vol. 9, pp. 1419-1433, 1966.

  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.

  9. Smolin, V.N. andPolyakov, V.K., Coolant Boiling Crisis in Rod Assemblies, Proc. of 6th Int. Heat Transfer Conf., Toronto, ON, Canada, pp. 47-52, 1978.


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