Development and validation of depletion code system IMPC-Burnup for ADS
Depletion calculation is important for studying the transmutation efficiency of minor actinides and long-life fission products in accelerator-driven subcritical reactor system (ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEA-ADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover, the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.
KeywordsADS-coupled proton–neutron transport Burnup calculation IMPC-Burnup FLUKA OpenMC ORIGEN2
The first author would like to acknowledge Mr. Hong Shuang for the technical support of FLUKA and OpenMC coupled calculation.
- 3.R.L. Moore, B.G. Schnitzler, C.A. Wemple, et al., MOCUP:MCNP-ORIGEN2 coupled utility program. INEL-95/0523 (1995)Google Scholar
- 4.G.L. Yu, K. Wang, Y.H. Wang, MCBurn—a coupling package of program MCNP and ORIGEN. Atom. Energy Sci. Technol. 37, 250–254 (2003). https://doi.org/10.3969/j.issn.1000-6931.2003.03.014. (in Chinese) CrossRefGoogle Scholar
- 6.A. Stankovskiy, G.V.D. Eynde, P. Baeten. et al., ALEPH2—a general purpose Monte Carlo depletion code. Paper presented at PHYSOR 2012: Conference on Advances in Reactor Physics—Linking Research, Industry, and Education. Knoxville, Tennessee, USA, 15–20 April (2012)Google Scholar
- 12.D. Pelowitz. MCNPX User’s Manual 2.7. 0. Los Alamos National Laboratory, Los Alamos, New Mexico (2011)Google Scholar
- 13.H.Q. Li, Y.W. Yang, The Development of Burn-up Program COUPLE by Combining MCNP and ORIGEN, Conference of Chinese Reactor Physics, Qinshan, China, Augest 8–12 (2004)Google Scholar
- 14.A.G. Croff, A User’s Manual for the ORIGEN2 Computer Code. ORNL/TM-7175 (1980)Google Scholar
- 20.M.D. Dehart, M.C. Brady, C.V. Parks, OECD/NEA burnup credit calculational criticality benchmark Phase I-B results. ORNL-6901 (1996)Google Scholar
- 21.I. Slessarev, A. Tchistiakov, IAEA ADS-benchmark results and analysis (TCM-Meeting, Madrid, Spain, Madrid, 1997)Google Scholar