Comparison of the nuclear data libraries in the shielding calculation for the accelerator facility of the Proton Engineering Frontier Project in Korea

Abstract

The radiation shielding calculations were performed for the beam experiment hall of the Proton Engineering Frontier Project (PEFP) in Korea using MCNPX code with nuclear data libraries of the ENDF-B/VII, JEFF3.1, LA150, JENDL-HE, and physics models, Bertini and Isabel built in MCNPX code. The neutron energy spectra from 6 targets (20 MeV/ 1.5 mA-Si, Cu, target, 20 MeV/ 50 uA-Si, 20 MeV/ 4.8 mA-C target, 100 MeV/ 300 uA-Si, W target, 100 MeV/ 10 uA-Si, 100 MeV/ 1.6 mA-C target) were calculated and used as radiation source terms. With these results, the concrete thickness satisfying shielding regulations in Korea was calculated for the target rooms and the building of the Beam Experiment Hall. Bertini, Isabel, JENDL/HE and LA150 nuclear data were applied to the evaluation of the proton induced neutrons. ENDF/B-VII, JEFF3.1, and JENDL/HE nuclear data were applied to the evaluation of the concrete shield. A transport method or physics, realization of the geometrical model, and nuclear data are important factors which decide the accuracy of the radiation transport. Of these, transport method or physics should be decided when the radiation transport code or methods is selected and a geometrical model has the limit of a description. Therefore, a selection of the nuclear data which will be used in the simulation is very important for the reliance of the radiation transport simulation. There are many nuclear data available in the radiation transport simulation (1). In this paper, ENDF/B-VII, JENDL/HE, and JEFF3.1 were compared in the aspect of the radiation shielding calculation. The physics models such as Bertini and Isabel were addi- tionally considered. MCNPX 2.5 code (2) using Monte Carlo Method was used in this study to compare nuclear data. A beam experiment hall in the proton accelerator facility with 2 MW under Proton Engineering Frontier Project (PEFP) in Korea was considered as a simulation model. 2 Benchmark calculation with MCNPX code The neutron source term and shielding effect of concrete are mainly considered in the accelerator shielding calculation. From these reasons, neutron production rate from the various targets and depth dose in the concrete wall were evaluated with various nuclear data and physics models.