|
ßäåðíà ô³çèêà òà åíåðãåòèêà
ISSN:
1818-331X (Print), 2074-0565 (Online) |
| Home page | About |
Development of neutron and photoatomic data libraries for the MCPV and MCSS codes
V. V. Ilkovych*
Institute for Nuclear Research, National Academy of Sciences of Ukraine, Kyiv, Ukraine
*Corresponding author. E-mail address:
vistaldi@gmail.com
Abstract: The paper presents experience gained in the development and verification of neutron and photoatomic libraries for the specialized computational codes MCPV and MCSS, which are used for the analysis of radiation exposure of VVER reactor pressure vessels, irradiation parameters of reactor internals, and irradiation conditions of surveillance specimens. Approaches to the generation of multigroup neutron libraries with different nuclide compositions and energy coverage, as well as neutron and photoatomic libraries in continuous-energy representation in the ACE format, are discussed. The verification of the developed libraries was performed by comparing calculation results obtained using different Monte Carlo codes in benchmark problems of neutron and photon transport in a simple spherical geometry.
Keywords: reactor dosimetry, neutron and photoatomic data libraries, Monte Carlo method, MCPV, MCSS.
References:1. Standard of organization of Ukraine SOU 73.1-23724640-001-2020. Quality System. Determination of Irradiation Conditions and Radiation Exposure of the VVER-1000 Reactor Pressure Vessel (Kyiv, Institute for Nuclear Research, NAS of Ukraine, 2020) 37 p. (Ukr)
2. Standard of organization of Ukraine SOU 73.1-23724640-002-2022. Quality System. Dosimetry of Surveillance Specimens of VVER-1000 Reactor Pressure Vessel Metal (Kyiv, Institute for Nuclear Research, NAS of Ukraine, 2022) 47 p. (Ukr)
3. V.N. Bukanov et al. MCPV Program Package for Calculation of Neutron Flux Functionals Impacting the VVER-1000 Reactor Pressure Vessel. Preprint of the Institute for Nuclear Research of the National Academy of Sciences of Ukraine KINR 05-6 (Kyiv, 2005) 28 p. (Rus)
4. J.F. Briesmeister (Ed.). MCNPTM – A General Monte Carlo N-Particle Transport Code. Version 4C. No. LA 13709-M (Los Alamos, NM, Los Alamos National Laboratory, 2000) 788 p. https://mcnp.lanl.gov/pdf_files/TechReport_2000_LANL_LA-13709-M_Briesmeisterothers.pdf
5. J. Leppänen et al. The Serpent Monte Carlo code: Status, development and applications in 2013. Ann. Nucl. Energy 82 (2015) 142. https://doi.org/10.1016/j.anucene.2014.08.024
6. R.E. MacFarlane et al. The NJOY Nuclear Data Processing System. Version 2016. LA-UR-17-20093 (Los Alamos, NM, Los Alamos National Laboratory, 2016) 802 p. https://usermanual.wiki/Document/LANL20182020NJOY20Nuclear20Data20Processing20System20Manual.1337592173.pdf
7. V.V. Ilkovych et al. Updating of data on neutron sources for MCPV and MCSS codes. In: XXV Annual Scientific Conference of the Institute for Nuclear Research of the NAS of Ukraine, Kyiv, Ukraine, April 16–20, 2018. Book of abstracts (Kyiv: INR NAS of Ukraine, 2018) p. 98. (Ukr) https://kinr.kyiv.ua/Annual_Conferences/KINR2018/pdf/book%20of%20%20abstracts_2018.pdf
8. J.M. Risner et al. Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. NUREG/CR-7045. ORNL/TM-2011/12 (Oak Ridge, TN, Oak Ridge National Laboratory, 2011) 108 p. https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7045/index
9. J.M. Risner et al. Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries. In: Reactor Dosimetry: 14th International Symposium. ASTM STP 1550, Bretton Woods, NH, USA, May 22–27, 2011 (West Conshohocken, PA: ASTM International, 2012) p. 561. https://doi.org/10.1520/STP155020120043
10. W. Haeck, B. Verboomen, E.M. Mbala. A Validated MCNP(X) Cross-Section Library Based on JEFF-3.1. SCK-CEN Reports. BLG-1034 (Mol, SCK•CEN, 2006) 126 p. https://researchportal.sckcen.be/en/publications/a-validated-mcnpx-cross-section-library-based-on-jeff-31/?utm_source=chatgpt.com
11. P.K. Romano et al. OpenMC: A State-of-the-Art Monte Carlo code for research and development. Ann. Nucl. Energy 82 (2015) 90. https://doi.org/10.1016/j.anucene.2014.07.048