Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

Document Type: Research Paper

Authors

Amirkabir University of Technology (Tehran Polytechnic), Department of Energy Engineering and Physics, Tehran, Iran

Abstract

Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core. 

Keywords

Main Subjects

[1] RELAP5 code development Team, “RELAP5/MOD3 Code Manual”, Idaho National Engineering Laboratory, 1995.

[2] Vahman, N., Akbari-Jeyhouni, R., Rezaei Ochbelagh, D., Amrollahi, R., “An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5/MOD3.2 and WIMSD-5B/PARCSv2.7”. Prog. Nucl. Energy 93, pp. 155-164, 2016.

[3] Pesaran, F., Jahanfarnia, G., Jafari, J., Abbaspour Tehrani-Fard, A., Mansouri, M., “Modeling of control rod ejection transient for VVER-1000-model 446 using RELAP5m3.3/PARCSv2.6 coupled codes”. Ann. Nucl. Energy 65, pp. 411-420, 2014.

[4] Andreeva, M., Pavlova, M.P., Groudev, P.P., “Investigation of critical safety function ‘‘Heat sink’’ at low power and cold condition for Kozloduy Nuclear Power Plant VVER- 1000/V320”. Ann. Nucl. Energy 40, pp. 221-228, 2012.

[5] Fernandez-Moguel, L., Birchley, J.,“Analysis of the accident in the Fukushima Daiichi nuclear power station Unit 3 with MELCOR_2.1”. Ann. Nucl. Energy 83, pp.193-215, 2015.

[6] Todreas, N.E., Kazimi, M.S., “Nuclear System, Thermal Hydraulic Fundamentals, Hemisphere Publishing Corporation”, New York, 1993.

[7] Calza-Bini, A., Cosoli, G., Filacchioni, G., Lanchi, M., Nobili, A., Pesce, E., Rocca, U., Rotoloni, P.L.,“In-pile measurement of fuel cladding conductance for pelleted and vipac zircaloy-2 sheathed fuel pin”. Nucl. Technol. 25, 103, 1975.

[8] Atomic energy organization of Iran (AEOI), “Final Safety Analysis Report (FSAR) for Bushehr WWER-1000 reactor”, Tehran, Iran, 2003.