MCNP-4A Code Research Articles

Design and evaluation of a TNA explosive-detection system to screen carry-on luggage

Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak 252Cf neutron source (1.03·107 n/s) and two 3"×3" NaI(Tl) detectors. The presence of explosives is confirmed via detection of the 10.83 MeV prompt gamma-ray associated with nitrogen decay. The MCNP4A code was used to simulate the neutron and gamma transport through the system. The thermal neutron flux in the activation position was measured using gold and indium foils. The measured thermal neutron flux was lower, by not more than 9.5%, than that of simulation. In this report the results of the preliminary tests on the system are described.

Measurements and calculations of neutron spectra modified by iron slabs bombarded by neutrons with energies up to 14 MeV

Neutron spectra behind iron slabs of different thicknesses have been measured in the energy range from 1.6 to 14 MeV using a proton recoil spectrometer and 2H( d, n) 3He and 9Be( d, n) 10B neutron sources. The measured results have been compared with the predicted ones using the three dimensional Monte Carlo code MCNP 4A and point-wise cross sections from the ENDF/B-IV and ENDF/B-VI data files. The results show that the ENDF/B-IV calculations are in better agreement with the experiment than those obtained by using the ENDF/B-VI cross section data.

Dosimetry errors in endovascular high-dose-rate brachytherapy

Monte Carlo data were used to demonstrate the dosimetry of the microSelectron high-dose-rate (HDR) iridium 192 ( 192Ir) stepping source. These data were used to assess the accuracy of the Nucletron brachytherapy planning system (BPS version 13) for peripheral vessel endovascular brachytherapy. Dose rates from the high-dose-rate (HDR) source are calculated using the Monte Carlo code MCNP4A. Calculations are made at 0.25-cm intervals in the longitudinal direction on sleeves of radii of 1 and 0.25 cm. The Monte Carlo data are summed and weighted to simulate the longitudinal dose distribution at a distance of 1 and 0.25 cm from an 192Ir source stepping through a straight pathway. A comparison is made between the simulated Monte Carlo dosimetry and the Nucletron brachytherapy planning system’s dosimetry. This study illustrates and quantifies the dosimetric errors at small distances associated with a point source dose calculation algorithm. The effects of step size, dwell time optimization, and active length on the accuracy of BPS v.13 for HDR endovascular brachytherapy are demonstrated.

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.

Overview of straight duct streaming experiments for ITER

Two straight duct streaming experiments for international thermonuclear experimental reactor (ITER) were performed by using the intense deuterium–tritium (D–T) neutron source at Japan Atomic Energy Research Institute (JAERI) in order to investigate the influence of streaming through ducts on radiation dose and to validate the accuracy of design calculations. One was a small duct streaming experiment for diagnostic ports, the other was a large duct streaming experiment for large ports such as neutral beam injector (NBI) port. The ducts increased the neutron flux above 10 MeV by 10 6–10 7 times at the ends of the ducts, while the increase of neutron flux below 10 MeV and γ-ray by the ducts was less than a few hundred times. These experiments were analyzed by the MCNP-4A code with three nuclear data libraries: FENDL/E-2.0, -1.0 and JENDL Fusion File. Though the ducts made complicated neutron and γ-ray flux distributions in a cavity region, most of the calculated results agreed within ±40% with the experimental data. It was demonstrated that the MCNP calculations with the nuclear data libraries were accurate enough for predicting nuclear design parameters of ITER even for various duct geometries as far as they were modeled precisely.

Investigation of neutron and photon flux spectra in a streaming mock-up for ITER

A neutronics mock-up of the International Thermonuclear Experimental Reactor (ITER) inboard shielding system with streaming path for the fusion neutrons was investigated in order to validate the design. Neutron and photon flux spectra were measured at several positions in the assembly. The experimentally determined flux spectra are analysed with the ITER design tools Monte Carlo code MCNP-4A and Fusion Evaluated Nuclear Data Library FENDL-1 and FENDL-2. Helium production, radiation damage and photon heating rates, which are design parameters of the shielding system are derived from both, the measured and the calculated spectral fluxes.

Measurement of Activation Reaction Rate Distributions in a Lead Assembly Bombarded with 500-MeV Protons

Reaction rate distributions of various activation detectors such as the natNi(n,x)58Co, 197Au(n,2n)196Au, and 197Au(n,4n)194Au reactions were measured to study the production and the transport of spallation neutrons in a lead assembly bombarded with protons of 500 MeV. The measured data were analyzed with the nucleon-meson transport code NMTC/JAERI combined with the MCNP4A code using the nuclide production cross sections based on the JENDL Dosimetry File and those calculated with the ALICE-F code. It was found that the NMTC/JAERI-MCNP4A calculations agreed well with the experiments for the low-energy-threshold reaction of natNi(n,x)58Co. With the increase of threshold energy, however, the calculation underestimated the experiments, especially above 20 MeV. The reason for the disagreement can be attributed to the underestimation of the neutron yield in the tens of mega-electron-volt regions by the NMTC/JAERI code.

Measurement and Analysis of Leakage Neutron Spectra from Spherical Assemblies of Chromium, Manganese and Copper with 14 MeV Neutrons

In order to make benchmark validation of the nuclear data for Cr, Mn and Cu in existing modern evaluated nuclear data files, leakage neutron spectra in the energy range from 0.1 to 15 MeV from spherical sample piles were measured by a time-of-flight technique using intense 14 MeV neutron source, OKTAVIAN. To make a qualitative understanding of the dependence of the spectra on individual partial cross sections, a trial calculation was performed by Monte Carlo neutron transport code MCNP-4A along with several kinds of modified cross sections for Mn. The measured spectra were compared with those by theoretical calculation using MCNP-4A with the evaluated nuclear data files. The comparison was made by the spectrum shape and by the C/E values in four energy regions. The calculations for Cr, Mn and Cu using JENDL-3.2, JENDL Fusion File, FENDL-1.0 and EFF-2 give satisfactory predictions with a few exceptions. By checking the result of the trial calculation, it is considered that these disagreement is caused by improper evaluation either of inelastic scattering cross section values and/or energy distribution of secondary neutrons.

A Fast Neutron Fluence Reduction Strategy for Pressure Vessel Lifetime Extension

The feasibility of nuclear power plant lifetime extension was examined by the evaluation of fast neutron (>1 MeV) fluence at the reactor pressure vessel (RPV). A fluence reduction option, additional shields installation in outer core structures, was applied to the Kori Unit 1 reactor, and its fluence reduction effect was carefully analyzed. Full-scope explicit modeling of Monte Carlo simulation with MCNP4A code was employed in the fast neutron fluence calculations. An optimized design option was found from various choices in geometry and material for shield structure. It is expected that the magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV. On the basis of the criterion of pressurized thermal shock requirement of the RPV, it is predicted to obtain the additional life of 4.4 effective full power years. It was also investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were negligibly influenced by the installation of additional shield structure.

Shadow Shield Design Analysis for Repository Emplacement Drifts

This paper describes the design approach and associated shielding analysis for radiological protection of occupational personnel in the subsurface facility of a potential repository that could be located at Yucca Mountain in Nevada. The VA (viability assessment) reference repository design provides a sufficient number of emplacement drifts for a capacity of 70,000 MTU spent nuclear fuel and defense high-level waste packages. The use of a simple, cost-effective concrete shadow shield after each last-emplaced waste package in an emplacement drift substantially eliminates direct radiation to subsurface workers in accessible areas such as the main drifts. The scattered radiation field becomes the major concern for worker safety in the subsurface facility. Adequacy of personnel protection was verified in a detailed shielding analysis using the MCNP4A code. Results show that the radiation field in the accessible main drifts is 1 mrem/hr at the worker elevation, once waste emplacement is complete and the drift isolation doors are closed. Therefore, the concrete shadow shield at the end of each emplacement drift provides adequate shielding for protection of repository workers.

Measurement of the Neutron Spectrum in the 14MeV Neutron Driven Vanadium-Alloy Assembly by the Multifoil Activation Method

A neutronics benchmark experiment on vanadium-alloy was conducted by using the D-T neutron source facility of FNS/ JAERI. The neutron spectra in the vanadium-alloy assembly which was bombarded by D-T neutrons were measured by the multifoil activation - adjustment method. Reaction rates of 20 dosimetry reaction were measured. Neutron spectra were calculated by the MCNP-4A code with the JENDL Fusion File and FENDL/E-1.0 libraries. These calculated neutron spectra were used as initial guesses and adjusted by the SAND-II code with the measured reaction rates. Cross sections of the FENDL/A-2.0 library were used as detector response functions for the adjustment. The obtained adjusted spectra are consistent with the results of the on-line spectrum measurement. The evaluated cross section data of vanadium-alloy were validated on the comparisons between the calculated initial spectra and the adjusted spectra.

Shielding and Criticality Characterization of ALR8(SI) Plutonium Storage Containers

An examination of the shielding and criticality safety performance of the Pantex-designed ALR8(SI) container for storage of ex-weapons plutonium pits is described. The MCNP-4A code was used as the major analysis tool. An MCNP-4A model of an ALR8(SI) container with a “hypothetical” ex-weapons pit inside was formulated. The hypothetical pit consisted of a hollow sphere of plutonium encased in stainless steel. For the photon shielding analysis, data for the gamma rays emitted during the decay of the plutonium isotopes were taken from the LANL website, “Map of the Nuclides.” Dose rates due to the photons emitted by the hypothetical pit’s decay were calculated at various positions external to the ALR8(SI) container. Since Pu-241 may undergo significant decay during the possible period of storage, the effect of varying the amount of Am-241 in the hypothetical pit’s plutonium was also considered. The neutron dose rates exterior to the ALR8(SI) container (with a hypothetical pit inside) were determined using two MCNP-4A calculations. In the first, the effective neutron multiplication factor (keff) of the container was calculated. In the second, the results from the keff calculation were used to define a surface neutron source at the pit’s outer edge, and the neutron dose rates external to the container were then computed based on this neutron source. The effective neutron multiplication factor (keff) of the container was evaluated both for dry conditions and for when the container was flooded with water.

Transport, Sensitivity and Uncertainty Analysis of FNG 14 MeV Neutron Bulk Shield Experiment

Bulk Shield experiment representing a mock-up of the ITER inboard blanket and vacuum vessel was completed in 1997 at the 14 MeV Frascati Neutron Generator (FNG) at ENEA Frascati. The objective of the experiment was to verify the calculational methods and nuclear data files used in the design and shielding calculations for fusion reactor. The benchmark experiment was analysed using the discrete ordinates transport code DORT, and the sensitivity and uncertainty analysis code package SUSD3D developed within the EFF project. Neutron cross-sections taken from FENDL-1 and FENDL-2 evaluations were used in the transport calculation. The results were compared with the measured reaction rates and with those calculated by the Monte Carlo code MCNP4A using FENDL-1. Good agreement between MCNP and DORT results was observed. No significant difference between FENDL-1 and FENDL-2 results was found, FENDL-2 performing slightly better. The transport calculations were combined with the sensitivity and uncertainty analysis whose application has proven to be very useful in order to obtain information on the importance of various nuclear data in the neutron transport, to explain the discrepancies between the calculation and the measurement, or at least to discard some possible reasons for the discrepancy. The sensitivity analysis coupled with the covariance matrices can guide future nuclear data research and development.

Investigation on Prediction Capability of Nuclear Design Parameters for Gap Configuration in ITER through Analysis of the FNS Gap Streaming Experiment

As an R&D Task of shielding neutronics experiment under the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), streaming experiments with simulating a gap configuration formed by two neighboring blanket modules of ITER were carried out at the FNS facility. In this work, prediction capability of various nuclear design parameters was investigated through analysis of the experiments. The Monte Carlo transport calculation code MCNP-4A and the FENDL/E-1.0 and JENDL Fusion File cross section data libraries were used for the analysis with detailed modeling of the experimental conditions. As a result, all the measured quantities were reproduced within about ± 30 % by the calculations. It was concluded that these calculation tools were capable of predicting nuclear design parameters, such as helium production rates at connection legs of blanket modules to the back plate and nuclear responses in toroidal field coils, with uncertainty of ± 30% for the geometry where gap streaming effect was significant.

Measurement of Secondary Gamma-ray Skyshine and Groundshine from Intense 14 MeV Neutron Source Facility

Secondary gamma-ray skyshine and groundshine, including the direct contribution from the facility building, have been measured with an Hp-Ge detector and an Nal(Tl) detector at the Intense 14 MeV Neutron Source Facility OKTAVIAN of Osaka University, Japan. The mechanism of secondary gamma-rays propagation were analyzed with the measured spectrum with the Hp-Ge detector. The contribution of the skyshine was shown to be a continuum spectrum that was composed of mainly Compton scattered high energy secondary gamma-rays generated in the facility building created by (n,γ) reaction. The contribution of the groundshine considerably contained secondary gamma-rays generated by natSi (n,γ) reaction in soil, including the albedo contribution from the ground. And the total contribution contained capture gamma-rays from iron (Fe) and other nuclides. The measurements with the Nal(Tl) detector as well as the Hp-Ge detector were carried out to investigate the dependence of gamma-ray dose as a function of distance from the neutron source up to hundreds meters. Consequently, it was found that the dependence could be fitted with the function of const. exp(-r/λ)/rn, where n values were around 2 except for the skyshine (n~1). It was thus indicated that the contribution of the skyshine could be propagated farther downfield than the direct contribution from the facility. The measured ratios of the three contributions (skyshine, groundshine, and direct contributions) and the distance dependence in each path were shown to be in good agreement with calculated results by the Monte Carlo transport code MCNP-4A. And the total contributions for the two detectors of Nal(Tl) and Hp-Ge agree excellently with each other.

Measurement and Analysis of Leakage Neutron Spectra from Spherical Assemblies of Chromium, Manganese and Copper with 14 MeV Neutrons.

In order to make benchmark validation of the nuclear data for Cr, Mn and Cu in existing modern evaluated nuclear data files, leakage neutron spectra in the energy range from 0.1 to 15 MeV from spherical sample piles were measured by a time-of-flight technique using intense 14 MeV neutron source, OKTAVIAN. To make a qualitative understanding of the dependence of the spectra on individual partial cross sections, a trial calculation was performed by Monte Carlo neutron transport code MCNP-4A along with several kinds of modified cross sections for Mn. The measured spectra were compared with those by theoretical calculation using MCNP-4A with the evaluated nuclear data files. The comparison was made by the spectrum shape and by the C/E values in four energy regions. The calculations for Cr, Mn and Cu using JENDL-3.2, JENDL Fusion File, FENDL-1.0 and EFF-2 give satisfactory predictions with a few exceptions. By checking the result of the trial calculation, it is considered that these disagreement is caused by improper evaluation either of inelastic scattering cross section values and/or energy distribution of secondary neutrons.

Effect of air cavities on the dose delivered to the lung during high-dose brachytherapy.

In the treatment of lung cancer using the radiotherapy technique of intracavitary brachytherapy with an 192Ir source, the lung is normally assumed to be entirely composed of a homogeneous mass of soft tissue. The aim of this study is to investigate whether there is the possibility that the air cavities in the lung influence the dose delivered to the lung at a prescribed distance from the source. The Monte Carlo code MCNP-4A was used to model the dose delivered by both 192Ir and 198Au as a function of treatment medium, density and composition, photon energy, and distance from the source. The suitability of MCNP-4A for this study was tested by producing depth-dose profiles for photons in water and comparing these to calculated profiles produced using well-documented methods.

Analytical Calculation of the Average Dancoff Factor for a Fuel Kernel in a Pebble Bed High-Temperature Reactor

An analytical expression was derived for the average Dancoff factor of a fuel kernel (Cfk) in a pebble of a high-temperature gas-cooled reactor. This Dancoff factor accounts for the probability that a neutron escaping from a fuel kernel enters another fuel kernel, in the same pebble or in other pebbles, without colliding with a moderator nucleus in between. If the fuel zone of the pebble is thought to be of infinite dimensions, the Dancoff factor becomes equal to the so-called infinite-medium Dancoff factor . The has been determined by the evaluation of three existing analytical expressions and by two Monte Carlo calculations performed with the MCNP-4A code, for various coated-particle densities. The Dancoff factor Cfk can be written as times a correction factor. The latter has been calculated for different fuel zone radii and pebble shell thicknesses. For the standard pebble, Cfk as a function of the number of coated particles has been calculated both analytically and with MCNP. The results of both methods are in good agreement. The analytical calculation method is preferred because it consumes practically no CPU time and obviates the building of MCNP models.

A Dosemeter Based on Nuclear Track Etched Detectors for Thermal, Fast and High Energy Neutrons with Flat Response

The response of a neutron dosemeter based on plastic track detectors has been studied. The dosemeter (converter + detector) configurations used consist of a CR-39 detector 500 μm thick and a Makrofol converter 300 μm thick for fast neutrons (configuration 1) and 3 mm of air used as converter for thermal neutrons (configuration 2). The possibility of using Makrofol as a high energy neutron dosemeter has also been studied. In order to validate the results obtained from Monte Carlo simulations using the MCNP4A code, a set of irradiations by monoenergetic neutron beams has been performed at the Physikalisch-Technische Bundesanstalt (PTB), Gesellschaft fur Strahlen-Und Umweltforschung (GSF) and Paul Scherrer Institut (PSI) neutron irradiation facilities. Irradiations with realistic fields were performed at the Cadarache Nuclear Centre. An excellent agreement has been found between the simulated and the experimental values, not only for fast neutrons but also for thermal neutrons.

Study on photon sensitivity of silicon diodes related to materials used for shielding

Abstract Large area silicon diodes used in electronic neutron dosemeters have a significant over-response to X- and gamma-rays, highly non-linear at photon energies below 200 keV. This over-response to photons is proportional to the diode's active area and strongly affects the neutron sensitivity of such dosemeters. Since silicon diodes are sensitive to light and electromagnetic fields, most diode detector assemblies are provided with a shielding, sometimes also used as radiation filter. In this paper, the influence of materials covering the diode's active area is investigated using the MCNP-4A code by estimating the photon induced pulses in a typical silicon wafer (300 μm thickness and 1 cm diameter) when provided with a front case cover. There have been simulated small-size diode front covers made of several materials with low neutron interaction cross-sections like aluminium, TEFLON, iron and lead. The estimated number of induced pulses in the silicon wafer is calculated for each type of shielding at normal photon incidence for several photon energies from 9.8 keV up to 1.15 MeV and compared with that in a bare silicon wafer. The simulated pulse height spectra show the origin of the photon-induced pulses in silicon for each material used as protective cover: the photoelectric effect for low Z front case materials at low-energy incident photons (up to about 65 keV) and the Compton and build-up effects for high Z case materials at higher photon energies. A simple means to lower and flatten the photon response of silicon diodes over an extended X- and gamma rays energy range is proposed by designing a composed photon filter.