Application of MCPN nonanalog techniques for calculations of reaction rate measurements at the BFS facilities

3/28/2016 2016 - #02 Modelling processes at nuclear facilities

Andrianova O.N. Manturov G.N. Rozhikhin Ye.V.

https://doi.org/10.26583/npe.2016.2.07

The BFS fast critical assemblies of the JSC «SSC RF – Institute for Physics and Power Engineering» are a unique experimental base for measuring different reactor functionals required to clarify and justify expected evaluations of both reactor core and ex-core characteristics. For years the data obtained on BFS integral critical experiments have been widely used to test and improve neutron cross-sections libraries and transport codes used for applications on fast reactor core design studies. Critical experiments performed at the BFS facilities provide a way to carry out a large series of studies needed for refining neutron data, including, for instance, measurements of central reactivity coefficients (reactivity introduced by material samples of various sizes) which allows testing the resonance structure of neutron cross-sections, but these experiments have not often been used previously for neutron cross-section adjustments. Experiments carried out in different years on various BFS core configurations and compositions were used for testing of project values of fast reactor models, studying physical properties of fuel, structural, and other materials, that brought to light data on neutron cross-sections used for correcting and adjusting neutron data for the neutron data library and transport codes.

An extensive experimental program included criticality, measurements of central reaction rate ratios, and fission rate distributions. The results of these measurements were evaluated and accepted for use as criticality and reactor physics parameters benchmarks. Calculations of criticality, central reaction rate ratios, and fission rate distributions were performed using the MCNP Monte-Carlo code using different files of evaluated nuclear data. The calculations showed that it is necessary to use variance reduction techniques to get the desired uncertainty of reaction rate calculations. Using of different variance reduction techniques implemented in MCNP for calculations of local functionals in critical multiplicating systems with complex heterogeneous geometry is considered in the report. Calculational analysis of effectiveness of variance reduction techniques was performed by the example of calculations of the central reaction rates. Results of criticality, central reaction rate ratios, and fission rate distributions calculations with different files of evaluated nuclear data are presented.

References

1. Nikolaev M.N. Konstantnoe obespechenie raschjotov bystryh reaktorov. Put’ ksovremennomu sostojaniju i zadachi dal’nejshego razvitija [Nuclear data for calculations of fast reactors. Way to recent state and tasks for future development]. Izvestia Visshikh Uchebnikh Zavedeniy. Yadernaya Energetika. 2013, no 4, pp. 5-16 (in Russian).
2. Briesmeister J.F., Brown J.S. MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, User’s Guide, Volume II. LA-CP-03-0245, LANL, 2003. 484 p.
3. Peregudov A.A., Andrianova O.N., Manturov G.N., Raskach K.F., Semenov M. Yu., Tsibulya A.M. Ispolzovanie metoda GRS dlya ocenki pogreshnosti neitronno-fizicheskih harakteristik perspektivnogo bistrogo reaktora [GRS Method to Evaluate Uncertainties in Calculation Parameters of an Advanced Fast Reactor]. Izvestia Visshikh Uchebnikh Zavedeniy. Yadernaya Energetika. 2014, no 2, pp. 90-98 (in Russian).
4. Kochetkov A.L., Matveenko I.P., Semenov M. Yu., Khomiakov Yu.S., Tsiboulia A.M. Kriticheskie jeksperimenty na stende BFS-2 v podderzhku razmeshhenija MOX-topliva v reaktory BN [Critical experiments on BFS-2 facility within maintenance of placement MOX-fuel in reactors BN]. Izvestia Visshikh Uchebnikh Zavedeniy. Yadernaya Energetika. 2007, no 2, pp. 16-27 (in Russian).
5. Dulin V.A., Kazanskii Yu.A., Mamontov V.F., Sidorov G.I. Some Physical Investigations in BFS-1 Fast Critical Assemblies. Atomic Energy. 1976, v. 40, no 5, pp. 377 - 381 (in Russian).
6. Doulin V., Kochetkov A., Pavlova O., Semenov M., Rozhikhin Ye. BFS1-FUND-EXP-003. Experimental Program Performed at the BFS-42 Assembly - K-Infinity Experiments for 238U in Fast Neutron Spectra: Measurements with Plutonium Mixed with Depleted Uranium Dioxide and Polyethylene. International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2007)1, March 2007.
7. Doulin V., Kochetkov A., Pavlova O., Semenov M., Rozhikhin Ye. BFS2-FUND-EXP-001. Experimental Program Performed at the BFS-31 Assembly - K-Infinity Experiments for 238U in Fast Neutron Spectra: Measurements with Plutonium Mixed with Depleted Uranium Dioxide and Polyethylene. International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2007)1, March 2007.
8. Doulin V., Kochetkov A., Pavlova O., Semenov M., Rozhikhin Ye. BFS2-FUND-EXP-001. BFS-97, -99, -101 Assemblies: Experimental Program on Critical Assemblies with Heterogeneous Compositions of Plutonium, Depleted-Uranium Dioxide, and Polyethylene. International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2007)1, March 2007.
9. Doulin V., Kochetkov A., Pavlova O., Semenov M., Rozhikhin Ye. BFS-57 and BFS -59 Assemblies: Experimental Program on Critical Assemblies with Heterogeneous Compositions of Enriched-Uranium Dioxide or Plutonium, Depleted-Uranium Dioxide, and Polyethylene. International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/ DOC(2007)1, March 2007.
10. Hodgdon A.D. A Variance Reduction Management Algorithm for MCNP. Transactions of the ANS, New Orleans. November 16-20, 2003, v. 89, pp. 373-374, 2003.
11. Smith H.P., Wagner J.C. A Case Study in Manual and Automated Monte Carlo Variance Reduction with a Deep Penetration Reactor Shielding Problem. Nuclear Science and Engineering. 2005, v. 149. pp. 23–37.
12. Koscheev V.N., Manturov G.N., Nikolaev M.N., Tsiboulia A.M. Verifikacija nejtronnyh dannyh osnovnyh reaktornyh materialov iz biblioteki rosfond na integral’nyh jeks- perimentah [Verification of neutron data for main reactor materials from rosfond neutron data library on intergral experiments]. Izvestia Visshikh Uchebnikh Zavedeniy. Yadernaya Energetika. 2014, no 1, pp. 204 – 214 (in Russian).
13. Andrianov A.A., Korovin Yu.A., Kuptsov I.S., Stankovskiy A.Yu. Interactive Information System for Preparation and Verification of Nuclear Data in the High-Energy Range/ Journal of the Korean Physical Society. August 2011, v. 59, no. 23, pp. 1096-1099.
14. Andrianova O., Peregudov A., Raskach K., Tsibulya A. Application of GRS Method to Evaluation of Uncertainties of Calculation Parameters of Perspective Sodium-Cooled Fast Reactor // Proc. Int. Conf. on Reactor Physics PHYSOR 2012. Knoxville, TN, USA, April 15-20, 2012.
15. Andrianova O., Koscheev V., Lomakov G., Manturov G. Neutron data adjustment based on integral critical experiments on the BFS-facility with different neutron spectrum. PHYSOR 2016, Sun Valley, ID 16222 pp. 2166-2175, May 1–5, 2016.
16. Evaluation of Proposed Integral Critical Experiments with Low-moderated MOX Fuel, Report NEA OECD no 6047, ISBN 92–64–01049–1, 2005. 124 p.
17. Zabrodskaya S.V., Ignatyuk A.V., Koscheev V.N., Manohin V.N., Pronjaev V.G. ROSFOND – Rossijskaya nacional’naya biblioteka ocenennyh nejtronnyh dannyh [ROSFOND – The National Library of evaluated neutron data]. Voprosy atomnoj nauki i tehniki. Ser.: Yadernye konstanty. 2007, no 1-2, pp. 3-21 (in Russian).

integral critical experiments BFS critical assemblies reaction rate ratios variance reduction methods MCNP

Link for citing the article: Andrianova O.N., Manturov G.N., Rozhikhin Ye.V. Application of MCPN nonanalog techniques for calculations of reaction rate measurements at the BFS facilities. Izvestiya vuzov. Yadernaya Energetika. 2016, no. 2, pp. 71-80; DOI: https://doi.org/10.26583/npe.2016.2.07 (in Russian).