Monte Carlo MCNP Research Articles

00/00784 On a simplified system for steam production in nuclear power plants with pressurized water reactors

The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

Using MCNP Code for Neutron and Photon Skyshine Analysis

The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector’s dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field.

Verification of ITER shielding capability and FENDL data benchmarking through analysis of bulk shielding experiment on large SS316/water assembly bombarded with 14MeV neutrons

Abstract The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

The use of MNCP for neutronics calculations within large buildings of fusion facilities

The calculation of nuclear parameters within fusion facilities is complicated by the complex geometry and large size of the proposed buildings housing the reactors. These complications make it impossible to use a single model, or code, to calculate the transport of neutrons from the plasma out into the rest of the building. In this paper, coupling two calculational models is demonstrated in calculating the operational dose rates in ITER building. The neutron and gamma fluxes during operation are calculated from the plasma region out to the cryostat of the machine using a two-dimensional discrete ordinates model (the subject of a companion paper) whereas a Monte Carlo MCNP model is applied in the rest of the building. In using this coupling approach, numerous joint Probability Mass Functions (PMFs) for the different phase space variables are used and constituted a specially-written source subroutine that is linked to MCNP. Along with the problem of proper source sampling and characterization, the physical size of the building, in comparison to the tally region, drastically complicates the calculation. As a result of this, the use of non-analog techniques are needed to help in the transport of particles in regions far away from the source (which is the NBI duct in this case). The fact that good results were easily achieved in the NBI room where there is a direct line of sight to the plasma, but as the detectors are placed further away, the results degenerate, exemplifying the need to use variance reduction techniques. Various techniques in the MCNP calculations are applied and compared and their usefulness is discussed. It is shown that MCNP can be used, in a limited fashion, to focus on specific high importance regions within large buildings such as in ITER.

The new remcounter LB6411: measurement of neutron ambient dose equivalent H*(10) according to ICRP60 with high sensitivity

Since the International Commission on Radiological Protection has issued in publication ICRP60 new recommendations on radiation protection quantities, in neutron monitoring there is now increasing interest in commercially available instruments optimized and calibrated for the measurement of ambient dose equivalent H*(10). Therefore within a joint cooperation between the Research Center Karlsruhe and EG&G Berthold the neutron-dose-rate meter LB6411 was newly developed. The detector system with integrated electronics has a /sup 3/He proportional counter tube centered in a moderating sphere. The response between thermal energies and 20 MeV was optimized with the help of extensive MCNP Monte-Carlo calculations The instrument has extremely high sensitivity of approximately 3 counts per nSv and can be used both as a portable or as a stationary neutron monitor. Fluence responses and angular dependencies had been measured in monoenergetic neutron beams provided by the Physikalisch-Technische Bundesanstalt (PTB) in Braunshweig, Germany. The ambient dose equivalent response of the LB6411 is reported over the whole energy range.

Validation of MCNP for RBMK Criticality Calculations

The Los Alamos National Laboratory Monte Carlo MCNP code is applied to critical experiments performed at the RBMK critical facility of the Russian Research Center “Kurchatov Institute,” Moscow. The validation investigations are completed by whole-core criticality calculations of experiments at the Smolensk Unit 3 nuclear power plant as part of the start-up procedure. The geometric model exploits the powerful capabilities of MCNP by precise representation of the fuel and different types of nonfuel channels, which add up to a detailed model of the criticalfacility and the RBMK core. Continuous-energy cross-section tables are taken from the ENDF/B-IV and ENDF/B-VI libraries. As the most important uncertainty inherent to the experimental setup, the concentration of impurity isotopes in the graphite moderator is identified. Within the resulting error limits, the keffand the void effect are well reproduced with both cross-section libraries.

Measurement and Analysis of an Angular Neutron Flux on a Beryllium Slab Irradiated with Deuterium-Tritium Neutrons

Angular neutron fluxes leaking from the surface of beryllium slab assemblies have been measured with irradiation of deuterium-tritium neutrons. The experiment was performed using the time-of-flight technique with an NE-213 scintillation detector. The measured neutron energy range was from 50 keV to 15 MeV. The thicknesses of the slabs were 50.8 and 152.4 mm, and the measured angles of the angular fluxes were 0.0, 12.2, 24.9, 41.8, and 66.8 deg. The experimental results have been compared with the results calculated by the Monte Carlo codes, MORSE-DD and MCNP, using the data of beryllium in the JENDL-3PR1, ENDF/B-IV, and Los Alamos National Laboratory nuclear data files. The results calculated with these files showed discrepancies of 20 to 30% from the experimental results. It was pointed out that the angular distributions of an elastic cross section and the total cross section of an inelastic reaction for 14.8-MeV neutrons in the files were insufficient to reproduce the measured spectra.

Transport calculations of neutron energy spectra in a graphite cylinder with a D-T source

Neutron energy spectra resulting from the transport of 14.7 MeV neutrons from a collimated D-T source through a graphite cylinder, have been calculated with the discrete-ordinates 2-D transport dot 4.2 code, with multigroup cross-sections generated using the njoy code from the ENDF/B (IV & V) libraries. The results confirm the conclusion of Goldfeld et al. (1985), that energy spectra at mesh points close to the axis of the system, in front of the collimated beam, consist mainly of one-collision contributions of elastically or inelastically scattered neutrons. Investigation of the dependence of the calculated spectra on the order of truncation of the Legendre polynomials expansion of the flux and of the cross-sections (i.e. the order of scattering) leads to the following observations: 1. (a) the P 6 or P 7 approximations seem to be adequate enough for flux calculations, with less than 3% error, in spite of the high degree of the source and the cross-section's anisotropy; 2. (b) the calculation error is reduced significantly by increasing the order of scattering from P 4 to P 7, mainly in mesh points close to the axis and of those energies in which the anisotropy of the elastic and discrete level inelastic scattering processes is most pronounced. Finally, the dot 4.2 calculations are compared with Monte Carlo mcnp calculations; both calculated spectra are in a good agreement.

Streaming of 14-MeV Neutrons Through an Iron Duct—Comparison of Measured Neutron and Gamma-Ray Energy Spectra with Results Calculated Using the Monte Carlo MCNP Code

Neutron and gamma-ray energy spectra resulting from the streaming of 14 MeV neutrons through a 0.30-m-diameter duct (length-to-diameter ratio = 2.83) have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data and data calculated previously using a combination of discrete ordinates and Monte Carlo methods. Comparisons are made at twelve detector locations on and off the duct axis for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The neutron spectra calculated using MCNP agree with the measured data within approx. 5 to approx. 50%, depending on detector location and neutron energy. Agreement with the measured gamma-ray spectra is also within approx. 5 to approx. 50%. The spectra obtained with MCNP are also in favorable agreement with the previously calculated data and were obtained with less calculational effort.