Substantiation of Data Files of JEFF-3.1.2 for Safety Analysis of TRIGA Mark-II Reactor through the Scrutiny of Integral Parameter of Benchmark Lattices TRX and BAPL
American Journal of Modern Physics
Volume 5, Issue 5, September 2016, Pages: 135-141
Received: Aug. 4, 2016;
Accepted: Aug. 22, 2016;
Published: Sep. 10, 2016
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Md. Mominul Islam, Department of Physics, Hajee Mohammad Danesh Science & Technology University, Dinajpur, Bangladesh
Md. Mahbubul Haque, Materials Science Division, Atomic Energy Centre Dhaka, Bangladesh Atomic Energy Commission, Dhaka, Bangladesh
S. M. Azharul Islam, Department of Physics, Jahangirnagar University, Savar, Dhaka, Bangladesh
The aim of this analysis is to bear out the nuclear data files of JEFF-3.1.2 for theoretical safety analysis of a 3 MW TRIGA MARK-II research reactor is custom-made at AERE, Dhaka, Bangladesh through the study of integral parameters of benchmark lattice TRX and BAPL of thermal reactor. The basic evaluated nuclear data files of JEFF-3.1.2 are selected for TRIGA reactor and processed by using nuclear data processing code NJOY99.0. Different cross-sections of U-235 and U-238 are computed from the NJOY output of the evaluated nuclear data library. The 69 group cross-section library is engendered from the processed file for reactor code WIMSD-5B. From the generated 69 group cross-section library, the integral parameters of yardstick lattices TRX and BAPL are premeditated by using cell code WIMSD-5B. The calculated integral parameters are compared to the deliberated values as well as the consequences of Monte Carlo Code MCNP. From the assessment it is found that all the integral parameters are in good concurrence with some suspicions. Through benchmarking the integral parameters of TRX and BAPL lattices this analysis reflects the support to the evaluated nuclear data files of JEFF-3.1.2 for safety analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.
Md. Mominul Islam,
Md. Mahbubul Haque,
S. M. Azharul Islam,
Substantiation of Data Files of JEFF-3.1.2 for Safety Analysis of TRIGA Mark-II Reactor through the Scrutiny of Integral Parameter of Benchmark Lattices TRX and BAPL, American Journal of Modern Physics.
Vol. 5, No. 5,
2016, pp. 135-141.
M. B. Chadwick, et al., “ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology”, Nuclear Data Sheets, vol. 107, pp. 2931-3060, December, 2006.
O. Allaoui, et al., “Validation of ENDF/B-VII.0 nuclear data library for shielding calculations using the Monte Carlo method”, International Journal of Advanced Research, vol. 2, pp. 55-62, 2014.
Cross Sections Evaluation Working Group, “ENDF-6 Formats Manual”, BNL-90365-2009, Brookhaven National Laboratory, pp. 1-13, 2009.
R. E. MacFarlane and D. W. Muir, “NJOY99.0: Code system for producing point-wise and multi-group neutron and photon cross sections from ENDF/B”, RSICC Code Package PSR-480. Los Alamos National Laboratory, Los Alamos, New Mexico, USA, 1999.
A. J. Koning, et al., “Status of the JEFF nuclear data library”, Journal of the Korean Physical Society, vol. 59, pp. 1057-1062, August 2011.
K. Shibata, et al., “JENDL-4.0: A new library for nuclear science and engineering”, Journal of Nuclear Science and Technology, vol. 48, pp. 1–30, 2011.
Z. Youxiang, L. Tingjin, Z. Jingshang and L. Ping, “CENDL-3- Chinese evaluated nuclear data library, version 3”, Journal of Nuclear Science and Technology, vol. 39, pp. 37-39, 2002.
ENDF-B/V-VI: The US Evaluated Nuclear Data Library, BNL-NCS-60496, Brookhaven National Laboratory, 1993.
BROND-2. Library of recommended evaluated neutron data, VANT, Ser. Nucl. Const, N 2-3.13, 1991.
A. J. Koning et al., “The JEFF evaluated nuclear data project”, Proceedings of the International Conference on Nuclear Data for Science and Technology, ND2007, Nice, France, 22-27 April 2007.
A. Santamarina, D. Bernard and Y. Rugama, “Validation Results from JEF-2.2 to JEFF-3.1.1”, JEFF Report 22, 2009.
J. R. Askew, F. J. Fayers and P. B. Kemshell, “A general description of the lattice code WIMS”, Journal of the British Nuclear Energy Society, vol. 5, pp. 564, 1966.
T. Kulikowska, “WIMSD-5B: A neutronic code for standard lattice physics analysis”, Distributed by NEA Data Bank. Saclay, France, 1996.
F. Leszczynski, “Description of Wims Library Update Project (WLUP)”, 2002 International Meeting on Reduced Enrichment for Research and Test Reactors, Bariloche, Argentina, November 3-8, 2002.
J. Hardy, Jr. D. Klein and J. J. Volpe, “A study of physics parameters in several water-moderated lattices of slightly enriched and natural uranium”, Nuclear Science and Engineering, vol. 40, pp. 101-115, 1970.
R. L. Hellens and G. A. Price, “Reactor physics data for water-moderated lattices of slightly enriched uranium”, Reactor Technology Selected Reviews-1964, 529.
J. R Brown, D. R. Harris, F. S. Frantz, J. J Volpe, J. C. Andrews and B. H. Noordhoff, Kinetics and buckling measurements in lattices of slightly enriched U or UO2 rods in H2O. WAPD-176, January, 1958.
M. Halder and S. M. T. Islam, “Comparative study of generated wimsd-5b multi-group constants library based on JENDL-3.2 with JEFF-3.1.1 CENDL-3.0 and original WIMS and validation of generated library through some benchmark experiments analysis”, IOSR Journal of Applied Physics, vol. 8, pp. 39-43, 2016.
M. N. Uddin, M. M. Sarker, M. J. H. Khan and S. M. A. Islam, “Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors”, Annals of Nuclear Energy, vol. 36, pp. 1521-1526, 2009.
R. Sher and S. Fiarman, “Studies of Thermal Reactor Benchmark Data Interpretation: Experimental Corrections”, EPRI NP-209, US, October 1976.
Cross Section Evaluation Working Group (CSEWG), “Benchmark specifications with supplements”, Brookhaven National Laboratory, National Nuclear Data Center, Upton, New York 11973, BNL-19302, II. ENDF-202, USA, November, 1986.
K. Benaalilou, et al., “A comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA MARK-II research reactor at CNESTEN using the cross-section ENDB-VII and JEFF3.1”, International Journal of Current Research, vol. 6, pp. 4519-4523, 2014.