MODELLING OF THE RADIOACTIVE FISSION PRODUCTS BEHAVIOR DURING SEVERE ACCIDENT AT NPP WITH VVER, TAKING INTO ACCOUNT THE UNCERTAINTY OF INITIAL DATA

ISSUES

Authors & Affiliations

Shmelkov Yu.B.
National Research Center “Kurchatov Institute”

Shmelkov Yu.B. – Chief Engineer, Cand. Sci. (Tech.). – Deputy Head of Department, Dr. Sci. (Tech.). Contacts: 1, pl. Akademika Kurchatova, Moscow, Russia, 249033. Tel.: +7 (917) 547-90-62; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..

Abstract

The article presents the results of the fission products(FP) behavior modelling for the severe accident scenario “Double-ended guillotine rupture of the main circulation pipeline with simultaneous complete blackout” for two types of reactor plants: VVER-1000 and VVER-1200. Modelling was carried out using a certified MAVR-TA code, which has a module for the uncertainty analysis and sensitivity analyzing of sensitivity to uncertainty. Results on the FP release from fuel and melt, including the molten corium concrete interaction during the accident at VVER-1000, were obtained, as well as results on the FP transport in the containment and the FP emission into the environment.
Modelling of fission product behavior was accompanied by uncertainty analysis and sensitivity analysis. The uncertainty analysis was carried out using a methodology based on the ASME V&V 20 standard and RB-166-20. A total of two hundred calculations for each of the processes (FP release from the fuel, FP release from the melt, FP transport in the containment) were performed during the uncertainty analysis. The number of variable parameters ranged from 10 to 20.
The convergence of the results depending on the number of calculations was assessed, and it was shown that convergence is achieved at a maximum of 120 calculations. The uncertainty ranges for several isotopes have been determined: Kr-85, Cs-137, Ba-140. As part of the sensitivity analysis, influence of varied parameters on the result was assessed with Kendall's rank correlation coefficient considered as a criterion of influence. It has been shown that the fuel sample temperature and the melt surface temperature have the greatest influence on the FP release from the fuel and FP release from the melt, while the containment leakage value has the greatest influence on the FP release into the environment.

Keywords
fission products, release from fuel, release from melt, emission to the environment, uncertainty analysis, severe accident, VVER, MAVR-TA

Article Text (PDF, in Russian)

References

NP-001-15. Obshchiye polozheniya obespecheniya bezopasnosti atomnykh stantsiy [NP-001-15. General provisions for ensuring the safety of nuclear power plants]. Moscow, Federal'naya sluzhba po ekologicheskomu tekhnologicheskomu i atomnomu nadzoru, 2015. 75 p.
SSG-2 Deterministic safety analysis for Nuclear Power Plants. IAEA Specific Safety Guide. Vienna, Austria, 2010.
Shmelkov Yu.B., Zvonarev Yu.A., Shutov N.V., Petrov L.V. Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a VVER NPP. Part 1 – Modelling of the release of fission products from the fuel. Nuclear Engineering and Design, 2012, vol. 385, p. 111407. DOI: https://doi.org/10.1016/j.nucengdes.2021.111407.
Shmelkov Yu.B., Zvonarev Yu.A., Shutov N.V., Petrov L.V. Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a VVER NPP. Part 2 – Modelling of the fission products transport in the primary circuit and inside the containment. Nuclear Engineering and Design, 2021, vol. 382, p. 111377. DOI: https://doi.org/10.1016/j.nucengdes.2021.111377.
Shmelkov Yu.B., Shutov N.V., Zvonarev Yu.A., Petrov L.V. Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a VVER NPP. Part 3 – Modelling of the release of fission products from the melt. Nuclear Engineering and Design, 2023, vol. 415, p. 112694. DOI: https://doi.org/10.1016/j.nucengdes.2023.112694.
Bolshov L.A. Results of SOCRAT code development, validation and applications for NPP safety
assessment under severe accidents. Nuclear Engineering and Design, 2019, vol. 341, p. 326. DOI: http://dx.doi.org/10.1016/j.nucengdes.2018.11.013.
Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer – V&V 20. ASME, 2009. 101 p.
RB-166-20. Rekomendatsii po otsenke pogreshnostey i neopredelennostey rezul'tatov raschetnykh analizov bezopasnosti atomnykh stantsiy [RB-166-20. Recommendations for assessing errors and uncertainties in the results of calculation analyzes of the safety of nuclear power plants]. Moscow, Rostekhnadzor, 2020. 24 p.
Glaeser H. Summary of existing uncertainty methods. OESD/CSNI Workshop on Best Estimate Methods and Uncertainty Evaluations. NEA/CSNI/R(2013)8/PART2. 2011.

UDC 621.039.586

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2024, no. 2, 2:10

BROND-3.1

ISSUES

2024, Issue 2

MODELLING OF THE RADIOACTIVE FISSION PRODUCTS BEHAVIOR DURING SEVERE ACCIDENT AT NPP WITH VVER, TAKING INTO ACCOUNT THE UNCERTAINTY OF INITIAL DATA