Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

The solution of the neutron transport equation is the basic input for the reactor physics design of a nuclear reactor system. Because of the complexities of geometry and cross-section data, the neutron transport equation is generally solved using numerical methods. One of the difficulties in the solution using these methods concerns the accuracy of distribution of neutron flux in the system containing void regions in a highly absorbing medium. The difficulty arises because of the issue in the continuity of flux at the void and material interface. There is a sudden and large change in the flux at this interface. Kobayashi benchmarks are widely used problems for testing the ability and accuracy of reactor physics codes to compute neutron flux distribution on such systems. An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The Kobayashi benchmarks have been used to test the PATMOC code to verify its accuracy in flux computation for systems with voids in-between high absorbing materials. Our results show that the PATMOC estimated values of flux distribution in the system compare very well with the reference results provided with the benchmark. In this paper we present the detailed results and analyses of this benchmark with PATMOC code

Simulation of random medium of fuel pebble with liquid salt coolant using PebMC code

Pebble-bed High Temperature reactors (HTR) are proposed to generate heat for power production and process heat application such as Hydrogen (H2) production. A HTR fuel pebble consists of random distribution of several thousands of fuel particles, called Tri-Structural Isotropic (TRISO), in fuel zone followed by a graphite layer. This configuration along with coolant layer results in double heterogeneity effect. A multi-group Monte Carlo code PebMC has been developed to accurately analyse molten salt cooled HTR lattice cell. Neutron transport in such a complex fuel distribution requires resonance self-shielded cross-section which is generated using rational approximation and Dancoff correction. Dancoff factor accounts for those neutrons leaving a fuel particle and directly colliding with another fuel particle, without any collision with moderator nuclei.In this study, an analytical expression based on a two-region model developed for gas cooled HTRs is adopted to calculate Dancoff factor for pebble with salt coolant. The implementation of these analytical expressions with 172groups WIMS library in PebMC code have been discussed. Subsequently, paper describes neutron tracking in random medium of HTR lattice using Monte Carlo method. The influence of fuel salt on Dancoff factor for various packing of TRISO particles and subsequent effect on infinite multiplication factor (K-inf) calculation have been presented. In addition, effect of temperature of molten salt on Dancoff factor have been studied

Neutron capture cross section measurement of 238U at the CERN n_TOF facility in the energy region from 1 eV to 700 keV

The aim of this work is to provide a precise and accurate measurement of the 238U(n,γ) reaction cross section in the energy region from 1 eV to 700 keV. This reaction is of fundamental importance for the design calculations of nuclear reactors, governing the behavior of the reactor core. In particular, fast reactors, which are experiencing a growing interest for their ability to burn radioactive waste, operate in the high energy region of the neutron spectrum. In this energy region most recent evaluations disagree due to inconsistencies in the existing measurements of up to 15%. In addition, the assessment of nuclear data uncertainty performed for innovative reactor systems shows that the uncertainty in the radiative capture cross section of 238U should be further reduced to 1–3% in the energy region from 20 eV to 25 keV. To this purpose, addressed by the Nuclear Energy Agency as a priority nuclear data need, complementary experiments, one at the GELINA and two at the n_TOF facility, were proposed and carried out within the 7th Framework Project ANDES of the European Commission. The results of one of these 238U(n,γ) measurements performed at the n_TOF CERN facility are presented in this work. The γ-ray cascade following the radiative neutron capture has been detected exploiting a setup of two C6D6 liquid scintillators. Resonance parameters obtained from this work are on average in excellent agreement with the ones reported in evaluated libraries. In the unresolved resonance region, this work yields a cross section in agreement with evaluated libraries up to 80 keV, while for higher energies our results are significantly higher

Fission fragment angular distribution measurements of 235U and 238U at CERN n_TOF facility

Neutron-induced fission cross sections of 238U and 235U are used as standards in the fast neutron region up to 200 MeV. A high accuracy of the standards is relevant to experimentally determine other neutron reaction cross sections. Therefore, the detection effciency should be corrected by using the angular distribution of the fission fragments (FFAD), which are barely known above 20 MeV. In addition, the angular distribution of the fragments produced in the fission of highly excited and deformed nuclei is an important observable to investigate the nuclear fission process. In order to measure the FFAD of neutron-induced reactions, a fission detection setup based on parallel-plate avalanche counters (PPACs) has been developed and successfully used at the CERN-n_TOF facility. In this work, we present the preliminary results on the analysis of new 235U(n,f) and 238U(n,f) data in the extended energy range up to 200 MeV compared to the existing experimental data

High accuracy 234U(n,f) cross section in the resonance energy region

New results are presented of the 234U neutron-induced fission cross section, obtained with high accuracy in the resonance region by means of two methods using the 235U(n,f) as reference. The recent evaluation of the 235U(n,f) obtained with SAMMY by L. C. Leal et al. (these Proceedings), based on previous n_TOF data [1], has been used to calculate the 234U(n,f) cross section through the 234U/235U ratio, being here compared with the results obtained by using the n_TOF neutron flux

High precision measurement of the radiative capture cross section of 238U at the n_TOF CERN facility

The importance of improving the accuracy on the capture cross-section of 238U has been addressed by the Nuclear Energy Agency, since its uncertainty significantly affects the uncertainties of key design parameters for both fast and thermal nuclear reactors. Within the 7th framework programme ANDES of the European Commission three different measurements have been carried out with the aim of providing the 238U(n,γ) cross-section with an accuracy which varies from 1 to 5%, depending on the energy range. Hereby the final results of the measurement performed at the n_TOF CERN facility in a wide energy range from 1 eV to 700 keV will be presented

The Nuclear Astrophysics program at n_TOF (CERN)

An important experimental program on Nuclear Astrophysics is being carried out at the n_TOF since several years, in order to address the still open issues in stellar and primordial nucleosynthesis. Several neutron capture reactions relevant to s-process nucleosynthesis have been measured so far, some of which on important branching point radioisotopes. Furthermore, the construction of a second experimental area has recently opened the way to challenging measurements of (n, charged particle) reactions on isotopes of short half-life. The Nuclear Astrophysics program of the n_TOF Collaboration is here described, with emphasis on recent results relevant for stellar nucleosynthesis, stellar neutron sources and primordial nucleosynthesis