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[1] USNRC, Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, Washington, D.C., USA, 1989.
[2] USNRC, Compendium of ECCS Research for Realistic LOCA Analysis, NUREG-1230, Washington, D.C., USA, 1988.
[3] USNRC, Proposed Rule, Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria, (ADAMS Accession No. ML12283A174) Washington, D.C., March, 2014.
[4] KINS, MARS-KS Code Manual, Volume II: Input Requirements, KINS/RR-1282, Rev.1, 2016.
[5] Geelhood, K.J. & Lusher, W.G., FRAPCON3.5: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behaviour of Oxide Fuel Rods for High Burnup, NUREG/CR-7022, May 2014.
[6] USNRC, RELAP5/MOD3 Code Manual, NUREG/CR-5535, Washington D.C., 2001.
[7] Zhang, H., Szilard, R., Epiney, A., Parisi, C., Vaghetto, R., Vanni, A. & Neptune, K., Industry Application ECCS/LOCA Integrated Cladding/ Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model, INL/EXT-17-42461, Idaho Falls, USA, 2017.
[8] Bang, Y.S, et al, Multiple Fuel Rods Modeling for LBLOCA Analysis of APR1400 under High Burnup Condition, 27th International Conference of Nuclear Energy for New Europe, Portoroz, Slovenia, 2019.
[9] Powers, D.A. & Meyer, R.O., Cladding Swelling and Rupture Models for LOCA Analysis, NUREG-0630. April 1980.
[10] KHNP, Final Safety Analysis Report, Shinkori Units 3 and 4, KHNP, Seoul, Korea, 2015.
[11] KEPCO, N.F., The Nuclear Design Report for Shin-Kori Nuclear Power Plant Unit 3
Cycle 2, KNF-S3C2-18011, Rev.0 (Proprietary), Daejeon, Korea, 2018.
[12] American Nuclear Society, Decay Energy Release Rates Following Shutdown of Uranium Fueled Thermal Reactors, Draft ANS-5.1/N18.5, October 1973
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Open Access
Research article

Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition

Young Seok Bang,
Joosuk Lee
Korea Institute of Nuclear Safety
International Journal of Energy Production and Management
|
Volume 4, Issue 1, 2019
|
Pages 86-95
Received: N/A,
Revised: N/A,
Accepted: N/A,
Available online: N/A
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Abstract:

Since the revision of the requirements to consider the effect of fuel burnup on emergency core cooling system performance was proposed, flow blockage in reactor core has been one of the important issues in the thermal-hydraulic analysis of loss-of-coolant accident (loca). The present paper describes how much flow blockage would be expected following a large break loca based on the actual nuclear design data including the power and burnup of the fuel rods. a system thermal-hydraulic code, mars-ks, is used for calculation where the burnup specific data of the fuel rods is supported by a fuel performance code, fracon3. To recover the weakness of the system code in which the flow blockage under multiple rods configuration cannot be automatically simulated in hydraulic calculation, a special modelling scheme is developed and applied to the calculation. The effect of flow blockage on the thermal-hydraulic response of the reactor core is also discussed. To compensate for the uncertainty of the present flow blockage model, additional calculations are attempted for a wide range of the level of blockage.

Keywords: effect of fuel burnup, flow blockage in reactor core, hydraulic modelling of swelling and rupture of cladding, large break LOCA, MARS-KS code
Data Availability

The data used to support the findings of this study are available from the corresponding author upon request.

Conflicts of Interest

The authors declare that they have no conflicts of interest.

References
[1] USNRC, Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, Washington, D.C., USA, 1989.
[2] USNRC, Compendium of ECCS Research for Realistic LOCA Analysis, NUREG-1230, Washington, D.C., USA, 1988.
[3] USNRC, Proposed Rule, Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria, (ADAMS Accession No. ML12283A174) Washington, D.C., March, 2014.
[4] KINS, MARS-KS Code Manual, Volume II: Input Requirements, KINS/RR-1282, Rev.1, 2016.
[5] Geelhood, K.J. & Lusher, W.G., FRAPCON3.5: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behaviour of Oxide Fuel Rods for High Burnup, NUREG/CR-7022, May 2014.
[6] USNRC, RELAP5/MOD3 Code Manual, NUREG/CR-5535, Washington D.C., 2001.
[7] Zhang, H., Szilard, R., Epiney, A., Parisi, C., Vaghetto, R., Vanni, A. & Neptune, K., Industry Application ECCS/LOCA Integrated Cladding/ Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model, INL/EXT-17-42461, Idaho Falls, USA, 2017.
[8] Bang, Y.S, et al, Multiple Fuel Rods Modeling for LBLOCA Analysis of APR1400 under High Burnup Condition, 27th International Conference of Nuclear Energy for New Europe, Portoroz, Slovenia, 2019.
[9] Powers, D.A. & Meyer, R.O., Cladding Swelling and Rupture Models for LOCA Analysis, NUREG-0630. April 1980.
[10] KHNP, Final Safety Analysis Report, Shinkori Units 3 and 4, KHNP, Seoul, Korea, 2015.
[11] KEPCO, N.F., The Nuclear Design Report for Shin-Kori Nuclear Power Plant Unit 3
Cycle 2, KNF-S3C2-18011, Rev.0 (Proprietary), Daejeon, Korea, 2018.
[12] American Nuclear Society, Decay Energy Release Rates Following Shutdown of Uranium Fueled Thermal Reactors, Draft ANS-5.1/N18.5, October 1973
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Bang, Y. S. & Lee, J. (2019). Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition. Int. J. Energy Prod. Manag., 4(1), 86-95. https://doi.org/10.2495/EQ-V4-N1-86-95
Y. S. Bang and J. Lee, "Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition," Int. J. Energy Prod. Manag., vol. 4, no. 1, pp. 86-95, 2019. https://doi.org/10.2495/EQ-V4-N1-86-95
@research-article{Bang2019Thermal-HydraulicRO,
title={Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition},
author={Young Seok Bang and Joosuk Lee},
journal={International Journal of Energy Production and Management},
year={2019},
page={86-95},
doi={https://doi.org/10.2495/EQ-V4-N1-86-95}
}
Young Seok Bang, et al. "Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition." International Journal of Energy Production and Management, v 4, pp 86-95. doi: https://doi.org/10.2495/EQ-V4-N1-86-95
Young Seok Bang and Joosuk Lee. "Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition." International Journal of Energy Production and Management, 4, (2019): 86-95. doi: https://doi.org/10.2495/EQ-V4-N1-86-95
BANG Y S， LEE J. Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition[J]. International Journal of Energy Production and Management， 2019， 4(1)： 86-95. https://doi.org/10.2495/EQ-V4-N1-86-95