Technology Breakthrough by Heavy Water Board in Material Support to Indian Nuclear Power Programme

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

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The nuclear-power programme

Journal of the Institution of Electrical Engineers ◽

10.1049/jiee-3.1960.0243 ◽

1960 ◽

Vol 6 (68) ◽

pp. 471-472

Keyword(s):

Nuclear Power ◽

Nuclear Power Programme

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Decoupling Criteria for Multi-Connected Equipment

Journal of Pressure Vessel Technology ◽

10.1115/1.2841892 ◽

1998 ◽

Vol 120 (1) ◽

pp. 93-98 ◽

Cited By ~ 2

Author(s):

G. R. Reddy ◽

H. S. Kushwaha ◽

S. C. Mahajan ◽

K. Suzuki

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Internal Structure ◽

Heavy Water ◽

Nuclear Power ◽

Seismic Analysis ◽

Real Life ◽

Water Reactor ◽

Reactor Building ◽

Heavy Water Reactor

Generally, for the seismic analysis of nuclear power plant structures, requirement of coupling equipment is checked by applying USNRC decoupling criteria. This criteria is developed for the equipment connected to the structure at one location. In this paper, limitations of this criteria and modifications required for application to real life structures such as pressurized heavy water reactor building are discussed. In addition, the authors endeavor to present a decoupling model for multi-connected structure-equipment. The applicability of the model is demonstrated with pressurized heavy water reactor building internal structure and steam generator.

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