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Special Section Papers

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

[+] Author and Article Information
Thambiayah Nitheanandan

Canadian Nuclear Laboratories,
Chalk River, ON K0J 1J0, Canada
e-mail: thambiayah.nitheanandan@cnl.ca

X. Cao

School of Mechanical Engineering,
Shanghai Jiao Tong University,
Shanghai, China
e-mail: caoxuewu@sjtu.edu.cn

J.-H. Choi

International Atomic Energy Agency,
Vienna, Austria
e-mail: jhchoi2@kepco-enc.com

D. Dupleac

Power Engineering Department,
Politechnica University of Bucharest,
Bucharest, Romania
e-mail: danieldu@cne.pub.ro

D.-H. Kim

Korea Atomic Energy Research Institute,
Daejeon, South Korea
e-mail: dhkim8@kaeri.re.kr

H. G. Lele

Bhabha Atomic Research Centre,
Mumbai, India
e-mail: hglele@barc.gov.in

A. K. Nayak

Bhabha Atomic Research Centre,
Mumbai, India
e-mail: arunths@barc.gov.in

H. P. Rammohan

Nuclear Power Corporation of India Limited,
Mumbai, India
e-mail: hprammohan@npcil.co.in

1Corresponding author.

Manuscript received July 29, 2016; final manuscript received November 22, 2016; published online March 1, 2017. Assoc. Editor: Arun Nayak.

ASME J of Nuclear Rad Sci 3(2), 020903 (Mar 01, 2017) (11 pages) Paper No: NERS-16-1084; doi: 10.1115/1.4035726 History: Received July 29, 2016; Revised November 22, 2016

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

Copyright © 2017 by ASME
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References

Figures

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Fig. 1

Schematic of an HWR reactor assembly

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Fig. 2

Schematic of a CANDU primary heat transport system: 1—steam line leading to turbines, 2—pressurizer, 3—steam generators, 4—pumps, 5—inlet headers, 6—calandria vessel, 7—fuel channel, 8—moderator circulation pump, 9—moderator heat exchanger, and 10—fueling machines

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Fig. 3

A comparison of the calculated PHTS pressure at the reactor outlet header (ROH)

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Fig. 4

A comparison of the time when liquid relief valves in the PHTS open for the first time in loops 1 and 2

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Fig. 5

A comparison of the coolant mass inventory in the steam generator

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Fig. 6

A comparison of total coolant mass in the PHTS and pressurizer

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Fig. 7

The time SG dryout calculated by participants

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Fig. 8

The variation of moderator inventory

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Fig. 9

Comparison of time from the accident initiation until the moderator was completely boiled off

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Fig. 10

The UO2 mass in the intact core, showing the variation in the core disassembly process. The KAERI results include the zirconium in the channels.

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Fig. 11

The corium mass (UO2 + Zr + ZrO2) in the calandria vessel terminal debris bed

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Fig. 12

The calculated calandria vessel wall temperature

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Fig. 13

The transient containment pressure calculated by the participants

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Fig. 14

Timing of containment equipment airlock seal failure

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Fig. 15

Total amount of hydrogen generated during the accident

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Fig. 16

The amount (mass) of fission products calculated in the primary heat transport system

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Fig. 17

The amount of noble gas released to the containment

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Fig. 18

Total amount of noble gas released

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Fig. 19

A comparison of the different stages of the station blackout sequence with increasing complexity

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