171ade7c-d032-4fbf-89df-0b698377e48320210825055756998naun:naunmdt@crossref.orgMDT DepositInternational Journal of Mechanics1998-444810.46300/9104http://www.naun.org/cms.action?id=2828128202112820211510.46300/9104.2021.15https://www.naun.org/cms.action?id=23280Calculation of Collision Probability Matrix of Nuclear Fuel Cell as a Function of Neutron Energy Group Using Flat Flux ModelMohammad AliShafiiDepartment of Physics, Andalas University, Padang, IndonesiaDianFitriyaniDepartment of Physics, Andalas University, Padang, IndonesiaSeni H JTongkukutDepartment of Physics, Sam Ratulangi University, Manado, IndonesiaZakiSu’udNuclear and Biophysics Laboratory, Bandung Institute of Technology, Bandung, IndonesiaOne of the methods that widely used in solving neutron transport equations in the nuclear fuel cell is the collision probability (CP) method. The neutron transport is very important to solve because the neutron distribution is related to the reactor power distribution. The important thing in the CP method is the CP matrix calculation, better known as has an important role in determining the neutron flux distribution in the reactor core. This study uses a linear flat flux model in each cell region for each energy group with white boundary condition. Although the type of reactor used in this study is a fast reactor, the matrix calculation still carried out in fast and thermal group energy. The matrix depends on the number of mesh in each cell region. The matrix formed from the mesh distribution will produce a matrix for each energy group. Because the boundary condition of the system is assumed that there are no contributions neutron source from the outside, the sum of the matrix must be less than one. In general, the results of the calculations in this study are following the theory8252021825202112112613https://www.naun.org/main/NAUN/mechanics/2021/a262003-013(2021).pdf10.46300/9104.2021.15.13https://www.naun.org/main/NAUN/mechanics/2021/a262003-013(2021).pdf10.1080/00223131.2017.1417177S. Miwa, Y. Yamamoto, G. Chiba. Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident. Journal of Nuclear Science and Technology, 2018, Vol. 55, No. 6, pp. 575- 598. 10.1016/j.pnucene.2018.12.008E. Cervi, A. Cammi, E. Zio. A new approach for nuclear reactor analysis based on complex network theory. Progress in Nuclear Energy, 2019, Vol. 112, pp. 96-106. 10.1016/j.apradiso.2018.12.004M. Lahdour, T. El Bardouni, E. Chakir, K. Benaalilou, M. Mohammed, H. Bougueniz, H. El Yaakoubi, H. NTPERSN: a new package for solving the multigroup neutron transport equation in a slab geometry. Applied Radiation and Isotopes, 2019, Vol. 145, pp. 73-84. 10.1016/j.cpc.2021.107885T. Younkin, D. L. Green, A. B. Simpson, B. D. Wirth, B. D. GITR: An accelerated global scale particle tracking code for wall material erosion and redistribution in fusion relevant plasma–material interactions. Computer Physics Communications, 2021, Vol. 264, pp. 107885. 10.1016/j.cpc.2020.107332S. Choi, D. Lee. Three-dimensional method of characteristics/diamond-difference transport analysis method in STREAM for whole-core neutron transport calculation. Computer Physics Communications, 2021, Vol. 260, pp. 107332. 10.1080/00411450.2014.913184D. Litskevich, B. Merk. SP3 solution versus diffusion solution in pin-by-pin calculations and conclusions concerning advanced methods. Journal of Computational and Theoretical Transport, 2014, Vol. 43, No. 1-7, pp. 214-239. 10.1088/1742-6596/877/1/012013M. A. Shafii, Analysis of Neutron Fission Reaction Rate in the Nuclear Fuel Cell Using Collision Probability Method with Non Flat Flux Approach. In Journal of Physics: Conference Series, 2017, Vol. 877, No. 1, pp. 012013 10.1051/matecconf/201819702006M. A. Shafii, J. Usman, S. H. Tongkukut, A. G. Abdullah. The Pij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method. In MATEC Web of Conferences 2018, Vol. 197, pp. 02006 Z. Suud, M. A. Shafii, R. S. R. SNM. Fission Yield Calculation Method and its Effect in Nuclear Fuel Cell Homogenization Calculation. Indonesian Journal of Physics, 2009, Vol 20, No. 2, pp. 33-36. K. Okumura, T. Kugo, K. Kaneko, K. Thuchihashi, K. A comprehensive neutronics calculation code system. Japan, JAEA, 2006, pp. 4-26. 10.4236/wjnst.2015.53014H. A. Yousef, A. H. El-Farrash, A. A. Ela, Q. Merza. Measurement of radon exhalation rate in some building materials using nuclear track detectors. World Journal of Nuclear Science and Technology, 2015, Vol. 5, No 03, pp. 141. 10.1080/18811248.2006.9711176T. Hazama, G. Chiba, K. Sugino. Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses. Journal of nuclear science and technology, 2006, Vol. 43, No. 8, pp. 908-918. 10.1016/j.anucene.2013.06.011O. Safarzadeh, A. Minuchehr, A. S. Shirani, Heterogeneous reactor core transport technique using response matrix and collision probability methods. Annals of Nuclear Energy, 2013, Vol. 62, pp. 137-143. 10.1088/1674-1137/38/5/058201C. Zhen-Ping, Z. Hua-Qing, S. Guang-Yao, S. Jing, H. LiJuan, H. Li-Qin, W. Yi-Can, Preliminary study on CADbased method of characteristics for neutron transport calculation. Chinese Physics C, 2014, Vol. 38, No. 5, art. id. 058201. 10.13182/nse15-127S. Ray, S. B. Degweker, R. Rai, K. P. Singh, A collision probability and MOC-based lattice and burnup code for analysis of LWR fuel assemblies. Nuclear Science and Engineering, 2016, Vol. 184, No. 4, pp. 473-494. 10.13182/nse15-28D. Altiparmakov, R. Wiersma, The collision probability method in today’s computer environment. Nuclear Science and Engineering, 2016, Vol. 182, No. 4, pp. 395- 416. 10.1016/j.pnucene.2019.103219D. Litskevich, B. Merk, S. Atkinson, Verification of the current coupling collision probability method with orthogonal flux expansion for the case of single cell. Progress in Nuclear Energy, 2020, Vol. 120, art. id. 103219. 10.13182/nse42-01-23T. Boševski, An Improved Collision Probability Method for Thermal-Neutron-Flux Calculation in a Cylindrical Reactor Cell. Nuclear Science and Engineering, 1970, Vol. 42, No. 1, pp. 23-27. I. V. Kazachkov, M. Yu. Kamaev, V. M. Khutornyi R. Ya. Tomyak, Modeling of the Thermal Processes in Repository of the Waste Nuclear Fuel with Water for Assessment of Potentially Hazardous Situations, WSEAS Transactions on Heat and Mass Transfer, 2019, Vol. 14, No. 17, pp. 137-146. T. Watanabe, Numerical Simulation of Droplet Combustion using Volume-of-Fluid Method, WSEAS Transactions on Heat and Mass Transfer, 2019, Vol. 14, No. 4, pp. 38-44.