Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Transient parameters analysis of non-scrammed local melting accident scenarios using MCNPx and COBRA/EN

Document Type : Scientific Note

Authors

B. Salmassian,
A. Rabiee
M.R. Nematollahi
A. Pirouzmand

Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, P.O.BOX: 71964-84334, Shiraz - Iran

https://doi.org/10.24200/nst.2022.1394
Abstract
The risk assessment of local nuclear fuel melting in nuclear power plants (NPPs) due to its localized nature and the difficulty in accurate monitoring is of vital importance for safe operation of nuclear reactors. Experiences have shown that flow blockage accident without safety control rod axe man (SCRAM) can lead to local nuclear fuel melting which consequently affect the safety of NPPs. The purpose of this study is to analyze the transient states leading to local fuel melting based on ATWS-related events given in FSAR of the VVER-1000/V446 nuclear reactor including pump failure, local blockage, power level increasing and their combinations. In this work, a coupling framework is first developed based on the MCNPx and the COBRA/EN codes. After validation with available data, the results showed that despite 18% deviation from the mass flow rate reduction limitation, 470 kPa from the channel pressure drop limitation, and 204K from the clad temperature limitation in the most pessimistic situation, the reactor SCRAM does not occur. However, in these conditions (where SCRAM does not occur), 70% void fraction for 12 minitues is observed in some channels. Therefore, there may be dry spots and local melting of fuel in normal operational and ATWS conditions that need to be identified. According to the results, the occurrence of the void fraction above zero is locally expectant and an appropriate monitoring system should be used to identify weakness points of the system.

Highlights

1. H. Nakamura, Particle Modeling of Fuel Plate Melting during Coolant Flow Blockage in HFIR, (2014).

 

2. J. Zhang, S. Yu, Y. Yan, Fixed-time output feedback trajectory tracking control of marine surface vessels subject to unknown external disturbances and uncertainties, ISA Transactions, 93, 145-155 (2019).

 

3. E.E. Lewis, Nuclear power reactor safety, 7791.

 

4. R. Gharari, et al., Study of Flow Path Blockage Accident Around a Hot Fuel Rod, (2016).

 

5. H.N. Dehjurian, Analysis Of Conjugate Conduction-Convection Heat Transfer In Nuclear Reactor Fuel Assembly, In Mechanical. Shiraz University (2015).

 

6. Y. Guo, et al., Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor, In 2018 26th International Conference on Nuclear Engineering, ASMEDC (2018).

 

7. B.R. Sehgal, Nuclear safety in light water reactors: severe accident phenomenology, Academic Press (2011).

 

8. I. Khamis, SAMG-D: The IAEA Training Toolkit on The Development of Severe Accident Management Guidelines, (2017).

 

9. P. Constantin, C. Foias, Navier-stokes equations, University of Chicago Press (1988).

 

10. I. Kataoka, A. Serizawa, Basic equations of turbulence in gas-liquid two-phase flow, International Journal of Multiphase Flow, 15(5),  843-855 (1989).

 

11. L. Ammirabile, Studies on supercritical water reactor fuel assemblies using the subchannel code COBRA-EN, Nuclear Engineering and Design, 240(10), 3087-4903 (2010).

 

12. D. Reddy, S. Sreepada, A. Nahavandi, Two-Phase Friction Multiplier Correlation for High-Pressure Steam-Water Flow, EPRI NP-2522, Research Project, 813 (1982).

 

13. L.S. Waters, MCNPX user’s manual, Los Alamos National Laboratory, (2002).

 

14. (AEOI), A.e.o.o.I., Inspection and Control Systems (I&A): Chapter 7, in Final Safety Assurance Report 2007, Federal State Unitary Enterprise “Research, Design, and Engineering Survey Institute “Atomenergoproekt”: Moscow (2007).

 

15. S.S. Arshi, S. Mirvakili, F. Faghihi, Modified COBRA-EN code to investigate thermalhydraulic analysis of the Iranian VVER-1000 core, Progress in Nuclear Energy, 52(6), 589-595 (2010).

 

16. M. Avramova, et al., Uncertainty analysis of COBRA-TF void distribution predictions for the OECD/NRC BFBT Benchmark.

Keywords

Local melting accident
Un-scrammed ATWS
VVER1000
COBRA/En
MCNPx
1. H. Nakamura, Particle Modeling of Fuel Plate Melting during Coolant Flow Blockage in HFIR, (2014).
 
2. J. Zhang, S. Yu, Y. Yan, Fixed-time output feedback trajectory tracking control of marine surface vessels subject to unknown external disturbances and uncertainties, ISA Transactions, 93, 145-155 (2019).
 
3. E.E. Lewis, Nuclear power reactor safety, 7791.
 
4. R. Gharari, et al., Study of Flow Path Blockage Accident Around a Hot Fuel Rod, (2016).
 
5. H.N. Dehjurian, Analysis Of Conjugate Conduction-Convection Heat Transfer In Nuclear Reactor Fuel Assembly, In Mechanical. Shiraz University (2015).
 
6. Y. Guo, et al., Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor, In 2018 26th International Conference on Nuclear Engineering, ASMEDC (2018).
 
7. B.R. Sehgal, Nuclear safety in light water reactors: severe accident phenomenology, Academic Press (2011).
 
8. I. Khamis, SAMG-D: The IAEA Training Toolkit on The Development of Severe Accident Management Guidelines, (2017).
 
9. P. Constantin, C. Foias, Navier-stokes equations, University of Chicago Press (1988).
 
10. I. Kataoka, A. Serizawa, Basic equations of turbulence in gas-liquid two-phase flow, International Journal of Multiphase Flow, 15(5),  843-855 (1989).
 
11. L. Ammirabile, Studies on supercritical water reactor fuel assemblies using the subchannel code COBRA-EN, Nuclear Engineering and Design, 240(10), 3087-4903 (2010).
 
12. D. Reddy, S. Sreepada, A. Nahavandi, Two-Phase Friction Multiplier Correlation for High-Pressure Steam-Water Flow, EPRI NP-2522, Research Project, 813 (1982).
 
13. L.S. Waters, MCNPX user’s manual, Los Alamos National Laboratory, (2002).
 
14. (AEOI), A.e.o.o.I., Inspection and Control Systems (I&A): Chapter 7, in Final Safety Assurance Report 2007, Federal State Unitary Enterprise “Research, Design, and Engineering Survey Institute “Atomenergoproekt”: Moscow (2007).
 
15. S.S. Arshi, S. Mirvakili, F. Faghihi, Modified COBRA-EN code to investigate thermalhydraulic analysis of the Iranian VVER-1000 core, Progress in Nuclear Energy, 52(6), 589-595 (2010).
 
16. M. Avramova, et al., Uncertainty analysis of COBRA-TF void distribution predictions for the OECD/NRC BFBT Benchmark.

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