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Please use this identifier to cite or link to this item: http://repository.iitr.ac.in/handle/123456789/22797
Title: Validation of computer code ‘ATMIKA’ against RD-14M Small Break LOCA experiments
Authors: Dixit A.
Yadav S.K.
Kumar N.
Khan T.A.
Hajela S.
Singhal, Mukesh Kumar
Published in: Nuclear Engineering and Design
Abstract: The simulation of Small Break Loss-of-Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility is performed under the auspices of International Atomic Energy Agency (IAEA) as an International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the in-house developed system thermal–hydraulic neutronic computer code ‘ATMIKA’, extensively used to analyze postulated events in Indian PHWRs. RD-14M is an 11 MW, full-elevation-scaled extensively instrumented thermal hydraulic Canadian test facility, possessing most of the key components of a CANDU (CANada Deuterium Uranium) primary heat transport system (PHTS). The loop configuration is similar to figure-of-eight geometry of a typical CANDU circuit and it is intended to reproduce the important geometric features of a reactor PHTS and the appropriate operating conditions. ‘ATMIKA’ prediction and its comparison against SBLOCA experimental results are compared in this paper. A specific SBLOCA experiment ‘B9006’ is selected for the Computer code ‘ATMIKA’ predictions. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection (ECI) and represents most complete SBLOCA test conducted in RD-14M that includes all the phases of the transient (blow-down, high-pressure ECI, secondary pressure ramp (crash cool), refill, low pressure ECI, exponential pump ramp, and natural circulation). This simulation demonstrates that ‘ATMIKA’ is adequately capable of predicting the break discharge, PHTS depressurization, channel flow rate, channel voiding, fuel sheath temperatures and high pressure core injection flow through ECCS accumulator and initiation of low pressure ECI for test B9006. ‘ATMIKA’ predicted results are compared with experimental results and it is seen that predicted results for all phases of transient are in good agreement with experimental results. © 2017 Elsevier B.V.
Citation: Nuclear Engineering and Design, 323: 427-433
URI: https://doi.org/10.1016/j.nucengdes.2017.07.042
http://repository.iitr.ac.in/handle/123456789/22797
Issue Date: 2017
Publisher: Elsevier Ltd
Keywords: ATMIKA
CANDU
IAEA
ICSP
PHWR
RD-14M
SBLOCA
ISSN: 295493
Author Scopus IDs: 57195281180
57190738316
15750861600
15750864400
15750741600
15751557600
Author Affiliations: Dixit, A., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Yadav, S.K., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Kumar, N., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Khan, T.A., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Hajela, S., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Singhal, M., Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Directorate of Reactor Safety & Analysis, Anushakti Nagar, Mumbai, 400094, India
Corresponding Author: Dixit, A.; Nuclear Power Corporation of India Limited, Anushakti Nagar, India; email: adixit@npcil.co.in
Appears in Collections:Journal Publications [HRE]

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