EDN: TTQSVI
Authors & Affiliations
Dolzhenkov E.A., Tomashchik D.Yu., Ryzhov N.I.
Nuclear Safety Institute of the Russian academy of sciences, Moscow, Russia
Dolzhenkov E.A. – Researcher. Contacts: 115191, Moscow, st. Bolshaya Tulskaya, 52. Tel.: +7 (961) 461-63-31; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Tomashchik D.Yu. – Researcher.
Ryzhov N.I. – Researcher.
Abstract
Simulation of the reactor behavior under severe accident conditions is mainly reduced to numerical modeling of the sequence of processes and phenomena that characterize the reactor state at each stage of the accident. The emergency state of the reactor is preceded by its operation in the normal operation mode, as a result of which the pre-accident state of the reactor is one of the factors influencing the accident progression. In turn, the data on the pre-accident state of the reactor is the initial data for calculating the accident. Since the progression of the severe accident is accompanied by a number of complex interrelated processes and phenomena, large uncertainties in the calculation results arise during modeling. Taking into account these uncertainties and the comprehensive nature of the calculation, it is considered reasonable to use a simplified nuclide kinetics model as part of a severe accident code, representing an efficient compromise between the accuracy and run-time. To demonstrate the capabilities of a simplified nuclide kinetics model concerning the calculation of the nuclide composition of the fuel and the decay heat power, the results of model validation on data from the open database OECD/NEA SFCOMPO 2.0 are presented, as well as the results of cross-verification of the model on the data given in RB-093-20. It is shown that the calculation results are consistent with the reference data with acceptable accuracy, which makes it possible to use the model as part of a severe accident code.
Keywords
nuclear fuel, WWER, PWR, fuel burn-up, fission product, actinides, evaluated nuclear data library, microscopic interaction cross-section, irradiation history, decay heat power, verification, validation, SFCOMPO 2.0
Article Text (PDF, in Russian)
References
1. Dolganov K.S., Dolzhenkov E.A., Fokin A.L. et al. Applicability of the nuclide kinetics fast estimate model for severe accident codes. Annals of Nuclear Energy, 2022, vol. 167. DOI: https://doi.org/10.1016/j.anucene.2021.108858.
2. Rearden B.T. et al. SCALE Code System. ORNL/TM‑2005/39 Version 6.2.3, 2018.
3. Jaboulay J.‑C., Brun E., Hugot F.‑X. et al. Rigorous‑two‑steps scheme of TRIPOLI‑4 Monte Carlo code validation for shutdown dose rate calculation. EPJ Web Conf., 2017, vol. 153, p. 02008. DOI: https://doi.org/10.1051/epjconf/201715302008.
4. Tarasov V.I. BONUS 1.2 Package. Radionuclide Production in Thermal Reactors. User Guide, IBRAE RAN, NSI‑SARR‑137‑2002.
5. Avvakumov A.V., Kiselev A.E., Mitenkova E.F. et al. Verification of the BONUS module integrated into the SOCRAT code. Atomic Energy, 2009, vol. 106, pp. 316–325. DOI: https://doi.org/10.1007/s10512-009-9170-x.
6. Dolzhenkov E.A. et al. Usovershenstvovaniye modeli nuklidnoy kinetiki v sostave integral'nogo koda SOKRAT/V3 [Improvement of Nuclide Kinetics Model in SOCRAT/V3 code]. Sbornik trudov 11-y Mezhdunarodnoy nauchno-tekhnicheskoy konferentsii “Obespecheniye bezopasnosti AES s VVER” [Proc. of the 11th Conf. “Safety Assurance of NPP with WWER]. JSC OKB «GIDROPRESS», May 21–24, 2019. Available at: http://www.gidropress.podolsk.ru/files/proceedings/mntk2019/autorun/index-ru.htm (accessed 15.03.2022).
7. Michel‑Sendis F. et al. SFCOMPO 2.0: An OECD NEA Database of Spent Nuclear Fuel Isotopic Assays, Reactor Design Specifications, and Operating Data. Annals of Nuclear Energy, 2017, vol. 100, pp. 779–788. DOI: https://doi.org/10.1016/j.anucene.2017.07.022.
8. RB-093-20. Radiatsionnyye i teplofizicheskiye kharakteristiki otrabotavshego yadernogo topliva vodo-vodyanykh energeticheskikh reaktorov i reaktorov bol'shoy moshchnosti kanal'nykh [Radiation, Thermal and Physical Characteristics of Spent Nuclear Fuel from Water‑Moderated Power Reactors and High Power Channel‑Type Reactors”]. Available at: https://docs.secnrs.ru/documents/rbs/%D0%A0%D0%91-093-20/%D0%A0%D0%91-093-20.pdf (accessed 16.03.2022).
9. Seleznev E.F. Kinetika reaktorov na bystrykh neytronakh. Pod red. akad. RAN A.A. Sarkisova [Fast Breeder Reactor Kinetics. Ed. Acad. RAS A.A. Sarkisov]. Moscow, Nauka Publ., 2013, 239 p.
10. JANIS User’s Guide. OECD/NEA, 2004.
11. NJOY2016. Available at: https://njoy.github.io/NJOY2016/ (accessed 15.03.2022).
12. A SME V&V 20. Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer – V&V 20, ASME, 2009.
13. Dolzhenkov E.A. et al. Estimation of system code SOCRAT/V3 accuracy to simulate the heat transfer in a pool of volumetrically heat liquid on the basis of BAFOND experiments. Annals of Nuclear Energy, 2021, vol. 151, p. 107902. DOI: https://doi.org/10.1016/j.anucene.2020.107902.
UDC 621.039.526
Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2022, no. 4, 4:1
