Calculation analysis of experiments on passing neutrons through the reflector layers at the BFSs critical assemblies for complementing the lead-cooled fast reactor verification database

9/16/2020 2020 - #03 Modelling processes at nuclear facilities

Zherdev G.M. Andrianova O.N. Borovskaya J.V. Zhirnov A.P. Teplukhina E.S.

https://doi.org/10.26583/npe.2020.3.13

The article presents the results of efforts on complementing the lead-cooled fast reactor verification database by analyzing and revising previously performed calculation and experimental studies on passing neutrons through the layers of the steel reflector carried out in different years on the IPPE BFS critical facilities. In particular, the considered data can be used for the assessment of uncertainties in the energy release in the lead-cooled fast reactor steel reflector. It was analyzed that experiments at the BFS-66 assembly on modeling neutron and photon fluxes in fast reactor shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 assembles on modeling the neutrons and gamma rays transport through the layers from various materials of fast reactor reflectors. The information and data presented in previously published materials describing these experiments has been analyzed and supplemented with relevant data necessary for the preparation of detailed calculation models for precision neutronic codes. Based on the updated and refined data, detailed, precise neutronic calculation models with a detailed specification of the heterogeneous structure of the BFS facilities and experimental devices have been developed and calculations have been carried out confirming their operability. The calculations were performed using Monte Carlo neutronic codes (MCU-BR, MCNP, MMK-RF, MMK-ROCOCO) with the BNAB-RF, MDBBR50 and ROSFOND nuclear data libraries of the main neutronic characteristics measured at the BFS-66, -64, -80-2 critical assemblies. The developed calculation models of the considered integral experiments can be used to justify the designs of lead-cooled fast neutron reactors, to verify the neutronic codes and nuclear data, to evaluate uncertainties in reactor characteristics associated with nuclear data.

References

1. DICE – Database for the International Criticality Safety Benchmark Evaluation Project. Available at: https://www.oecd-nea.org/science/wpncs/icsbep/dice.html (accessed Apr 06, 2020).
2. ICSBEP – International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95). Available at: https://www.oecd-nea.org/science/wpncs/icsbep/handbook.html (accessed Apr 06, 2020).
3. IRPhEP – International Reactor Physics Experiments Evaluation Project. Available at: https://www.oecd-nea.org/science/wprs/irphe/ (accessed Apr 06, 2020).
4. RB-061-11.Provision on Verification and Review of Software in the Field of «Neutron and Physical Calculations». Available at: https://files.stroyinf.ru/Index2/1/4293796/4293796222.htm (accessed Apr 06, 2020) (in Russian).
5. Dulin V., Matveenko I., Rozhikhin E., Tsiboulia A.M. An Overview of the Experiments Performed at the BFS Facilities and Evaluated for the International Reactor Physics Experiment Evaluation Project. Nucl. Sci. and Eng., 2014, v. 178, no. 3.
6. Gomin E.A, Gurevich M.I., Maiorov L.V. Status of MCU. Programme and Book of Abstracts. Advanced Monte Carlo on Radiation Physics, Particle Transport Simulation and Applications, Monte Carlo 2000, October 23 – 26, 2000. Lisbon, Portugal, 2000, pp. 2003-2004.
7. Andrianova O., Manturov G., Rozhikhin Yu. Application of MCNP nonanalog techniques for calculations of reaction rate measurements at the BFS facilities. Izvestiya vuzov. Yadernaya energetika. 2016, no. 2, pp. 66-76 (in Russian).
8. Jerdev G.M., Kislitsyna T.S., Nikolaev M.N. ROCOCO System of Combined Neutron Constants – Current Status and Results of Testing Using Geometrical Module of the MMK Code. Izvestiya vuzov. Yadernaya Energetika. 2018, no. 2, pp. 47-57; DOI: https://doi.org/10.26583/npe.2018.2.05 (in Russian).
9. Andrianova O., Golovko Ye., Manturov G. Verification of the ROSFOND/ABBN Nuclear Data Based on the OECD/NEA Benchmark on Criticality Safety of MOX-Fueled Systems. Izvestiya vuzov. Yadernaya Energetika. 2018, no. 3, pp. 160-170; DOI: https://doi.org/10.26583/npe.2018.3.09 (in Russian).
10. Mitenkova E.F., Novikov N.V. Particulars of Neutron-Physical Calculations of Sodium-Cooled Fast Reactors with Mixed Oxide Fuel. Atomnaya Energiya. 2011, v. 109, iss. 5, pp. 309-320 (in Russian).
11. Danilychev A.V., Korobeynikova L.V., Seregin A.S., Stogov V.Yu., Chernyy V.A. International Analysis of Reactivity Coefficients for IAEA Benchmark Model (Hybrid Core of BN-600 Reactor). VANT. Ser. Yadernye Konstanty. 2002, v. 1-2, pp.89-108 (in Russian).
12. Kochetkov A.L., Matveenko I.P., Semenov M.Ju., Khomiakov Yu.S., Tsiboulia A.M. Critical Experiments on BFS-2 Facility within Maintenance of Placement MOX-Fuel in Reactors BN. Izvestia Vysshikh Uchebnykh Zawedeniy. Yadernaya Energetika. 2007, no. 2, pp. 16-27 (in Russian).
13. Blyskavka A.A., Moiseev A.V., Semenov M.Yu., Seregin A.S., Khomyakov Yu.S., Tsibulya A.M., Belov A.A., Seleznev E.F., Vasil’ev B.A., Farakshin M.R. Accuracy Analysis of BN-600 Power Density Calculations. Atomnaya Energiya. 2010, v. 108, iss. 2, pp. 1-6 (in Russian).
14. Mitenkova E.F., Novikov N.V. The Calculated Neutronic Characteristics Against Reconstructed Ones at (U-Pu) Experiments. Proc. of the PHYTRA2, Fez, Morocco. September 26–28, 2011, Paper 056.

integral experiments BFS Monte Carlo codes ROSFOND ABBN MCU-BR MMK-RF MMK-ROCOCO

Link for citing the article: Zherdev G.M., Andrianova O.N., Borovskaya J.V., Zhirnov A.P., Teplukhina E.S. Calculation analysis of experiments on passing neutrons through the reflector layers at the BFSs critical assemblies for complementing the lead-cooled fast reactor verification database. Izvestiya vuzov. Yadernaya Energetika. 2020, no. 3, pp. 125-135; DOI: https://doi.org/10.26583/npe.2020.3.13 (in Russian).