Research Papers

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

[+] Author and Article Information
Alexander Vasiliev

Nuclear Safety Institute (IBRAE),
B. Tulskaya 52,
Moscow 115191, Russia
e-mail: vasil@ibrae.ac.ru

Juri Stuckert

Karlsruhe Institute of Technology (KIT),
Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen, Karlsruhe 76344, Germany
e-mail: juri.stuckert@kit.edu

1Corresponding author.

Manuscript received March 16, 2015; final manuscript received September 29, 2015; published online February 29, 2016. Assoc. Editor: Leon Cizelj.

ASME J of Nuclear Rad Sci 2(2), 021024 (Feb 29, 2016) (7 pages) Paper No: NERS-15-1027; doi: 10.1115/1.4031815 History: Received March 16, 2015; Accepted September 29, 2015

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

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Stuckert, J., Große, M., Rössger, C., Steinbrück, M., and Walter, M., 2014, “Influence of the Temperature History on Secondary Hydriding and Mechanical Properties of Zircaloy-4 Claddings: An Analysis of the QUENCH-LOCA Bundle Tests,” Proceedings of International Conference on Nuclear Engineering ICONE22, Prague, July 7–11, American Society of Mechanical Engineers (ASME), New York, Paper No. ICONE22-30792.
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Fig. 1

Schematic representation of QUENCH test section facility

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Fig. 2

Cross section of QLOCA-1 test bundle (21 heated, 4 corner rods); consecutive numbers of rods are indicated

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Fig. 3

SOCRAT nodalization for QUENCH-LOCA-1 test

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Fig. 4

QUENCH-LOCA-1 characteristic temperature behavior; consecutive numbers of test phases are indicated

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Fig. 5

QUENCH-LOCA-1 electric power (total, core, inner and outer rings) history

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Fig. 6

QUENCH-LOCA-1 mass flow rates of steam, argon, and reflood water

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Fig. 7

QUENCH-LOCA-1: temperature at elevation 950 mm

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Fig. 8

QUENCH-LOCA-1: temperature at elevation 1050 mm

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Fig. 9

QUENCH-LOCA-1: temperature at elevation 850 mm

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Fig. 10

QUENCH-LOCA-1: temperature at elevation 550 mm

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Fig. 11

QUENCH-LOCA-1: temperature at elevation 350 mm

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Fig. 12

QUENCH-LOCA-1: temperature at elevation −50  mm

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Fig. 13

QUENCH-LOCA-1 calculation heat balance: 1, core electric power; 2, power transferred by gas; 3, heat flux to shroud; 4, chemical power

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Fig. 14

QUENCH-LOCA-1: calculated hydrogen generation rate and integral production

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Fig. 15

QUENCH-LOCA-1: post-test view of rod #9 from inner row

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Fig. 16

QUENCH-LOCA-1: post-test view of rod #16 from outer row

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Fig. 17

QUENCH-LOCA-1: the calculated pressure in rods placed in the corresponding radial rings with different initial pressures



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