Monte Carlo Neutron Photon Transport Research Articles

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

Criticality search of soluble boron iteration in MC code JMCT

The soluble boron is used to control the reactivity in normal operation and accident of pressurized water reactor. A criticality search of soluble boron iteration algorithm based on neutron balance is implemented in the Monte Carlo neutron-photon transport code JMCT. Firstly, the interpolation methods and our algorithm of soluble boron iteration are well introduced and compared. Then, the boron concentration of BEAVRS in hot zero power is calculated and compared to the measured data to validate the algorithm.

Evaluation of the radiation damage parameters of ODS steel alloys in the first wall of deuterium-tritium fusion-fission (hybrid) reactors

One of the most important parameters in the design of the fusion-fission hybrid reactor is the selection of the first wall material. Because the oxide dispersion-strengthened (ODS) steel alloys have high temperature oxidation, high radiation resistance, good hardness, and corrosion resistance properties, they are thought first wall candidate materials for fusion and fission applications. The objective of this paper is to determine the best radiation damage parameters of various experimental and commercial ODS steels (namely, 12Y1, 12YWT, 1DS, IDK, Eurofer97, MA956, MA957, and PM 2000). Neutron spectrum and average neutron energy throughout blanket, displacement per atom, hydrogen and helium production, nuclear heating, and tritium breeding ratio were calculated by using Monte Carlo methods with Monte Carlo neutron-photon transport code and nuclear libraries named as ENDF/B-VI and CLAW-IV. It is assumed that calculated reactor has been operated full power during a year and neutron wall load is 2.25 MW/m2 (1014 n/s). All investigated first wall materials should be replaced between 3.5 and 4 years. All investigated materials provide minimum required tritium breeding ratio value, and when considering all the calculations performed in this work, 1DS ODS steel is the most suitable first wall materials with respect to other investigated ODS steels.

Development of the Advanced Neutronic Analysis Model for the K-DEMO With MCNP Code

Neutronic assessment is the important foundation of Korean fusion demonstration reactor (K-DEMO) study. Previously, preliminary neutronic calculations using a simplified model have been performed [1] , and optimized structure of blanket module and shield material has been derived. Based on the previous calculation results, the advanced K-DEMO neutronic analysis model has been developed to perform more accurate and reliable neutronic assessments with the detailed configuration of vacuum vessel (VV), toroidal field (TF) coil, blanket, and shield. The advanced K-DEMO neutronic analysis model was imported into the integrated interface program by the help of Monte Carlo automatic modeling system to produce input files for the Monte Carlo neutron photon transport code analyses. The neutronic calculation on neutron wall loading (NWL) distribution, tritium breeding ratio (TBR), and nuclear heating in tokamak components was conducted for the newly developed advanced K-DEMO neutronic analysis model. In the NWL calculation, the peak was indicated at the inboard (IB) and the outboard (OB) blanket first wall with the value of 3.81 and 3.93 MW/ $\text{m}^{2}$ . The different types of models by changing gaps between the blanket modules and the thickness of separated borated steel shield blocks are calculated to check global TBR and the shield performance. The global TBR of 1.017 was achieved with the 49.7- and 76.2-cm-thick IB and OB blanket modules. The model with 1-cm-thick assembly gap between blanket modules, 22-cm-thick IB shield, 50-cm-thick OB shield, and 34-cm-thick divertor shield are considered as a reasonable shield model with the total nuclear heating in 16 TF coils and VV of 9 and 32 kW.

MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, <TEX>$B_1$</TEX> theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

Nuclear spectroscopy for in situ soil elemental analysis: Monte Carlo simulations

We developed a model to simulate a novel inelastic neutron scattering (INS) system for in situ non-destructive analysis of soil using standard Monte Carlo Neutron Photon (MCNP5a) transport code. The volumes from which 90%, 95%, and 99% of the total signal are detected were estimated to be 0.23m3, 0.37m3, and 0.79m3, respectively. Similarly, we assessed the instrument's sampling footprint and depths. In addition we discuss the impact of the carbon's depth distribution on sampled depth.

Assessment of Primordial Radionuclides in Pakistani Red Bricks and Associated Radiation Doses

Specific activity of primordial radionuclides and associated radiation hazards due to 40K, 226Ra, and 232Th have been measured in backed red brick samples, collected from five highly populated areas of the North West Frontier Province of Pakistan. For the detection, analysis and data acquisition, a high purity germanium detector was used. Associated external doses were calculated using a Monte Carlo neutron photon transport code. A theoretical model to determine the gamma dose rate at 1 m height from the floor, made of bricks, was employed for the calculation of mass attenuation coefficient and self-absorption in the floor for the gamma energies of these radionuclides and their progeny. Monte Carlo simulation shows that in this study the floor, having more than an effective thickness of 15 cm, contributes very little to the external gamma dose rate. The values of the external dose rate and annual effective dose are found to be much lower than the world average as well as from other countries of the world.

Update of ITER 3D basic neutronics model with MCAM

In order to conduct nuclear analyses on neutronics issues for the International Thermonuclear Experimental Reactor (ITER), a standard three-dimensional (3D) neutronics model of the ITER reactor has been developed. However, the model for the Monte Carlo neutron–photon transport code (MCNP) should be updated and refined due to the CAD design changes. Monte Carlo Automatic Modeling system (MCAM) as an interface code between CAD software and Monte Carlo particle transport computer code such as MCNP, is a tool of modeling and visualization, which can convert a CAD model to a neutronics model for MCNP and vice versa. This contribution presents the details of MCAM application to the update and refinement of the ITER 3D basic neutronics model containing blanket segmentation update and incorporation of fine structures of the inboard blankets. Based on the updated neutronics model, the neutron wall loading distribution, nuclear heating in the blanket modules and TF coil inboard legs are re-calculated.

Neutronic Comparison of Tritium-Breeding Performance of Candidate Tritium-Breeding Materials

Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, is one of the key issues of fusion power development. The tritium breeding performances of various tritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2 data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Li on Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBR are also calculated.

Neutronics Optimization of LiPb-He Dual-Cooled Fuel Breeding Blanket for the Fusion-Driven Sub-Critical System

The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR >1.05) and annual output of 100 kg or more fissile 239Pu (FBR >0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimizated calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (BR = TBR + FBR) is listed corresponding to different cases.

Neutronics Optimization of Tritium Breeding Blanket for the FDS

Neutronics optimization calculations have been performed for the tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectic breeder Li17Pb83, respectively, based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

Refined Monte Carlo modelling of fusion neutron emission in magnetically confined plasmas

A new refined modelling procedure for calculating a series of weighted samples for birth distributions of fusion products from arbitrary ion velocity distributions is presented. The weighted samples can be used directly as input to other codes. This is particularly useful if the fusion products are neutrons and their migration from the plasma to a detector or a certain point in space is simulated by using the Monte Carlo neutron-photon transport code MCNP. As an application of the new sampling procedure, numerical results for a test plasma using data of a neutral-beam-heated discharge of the stellarator W7-AS are presented.

Measurements and simulations of the neutron production at W7-AS

The 2.5 MeV neutron diagnostic system on the W7-AS stellarator comprises a set of BF3 and He3 neutron counters, a NE-213 scintillator and activation foils. The neutron counter set provides a dynamic range of several orders of magnitude while the scintillator provides fast resolution needed for studying transient events. The counters are calibrated in situ using a Pu/B238 neutron source placed at discrete points inside the vacuum vessel. The activation samples have been used to determine the total neutron yields as well as to measure the fluence of fast and thermal neutrons at various positions in the diagnostic hall of W7-AS. Neutron transport calculations have been performed using the Monte Carlo neutron photon transport code in order to determine the neutron fluence and energy distribution at the irradiation positions and to simulate the measured counter signals for the in situ calibration. The accuracy of the neutronics simulations depends on the quality of the geometric model and the material composition of major structural components.

Special absorber neutron detector moderator assembly: A new detector system for flux measurements of collimated 2.5 MeV neutrons

A new detector system has been developed for time resolved flux measurements of collimated D-D neutrons in large magnetic confinement fusion experiments. The novel detection concept, called the special absorber neutron detector moderator assembly, is to place a scatterer behind a collimator and to detect the scattered neutrons in a small Bonner sphere. The system has been optimized by means of the Monte Carlo neutron photon transport code (MCNP) simulations and is capable to be extended for a neutron profile camera. First tests and calibration measurements of a system using a 3 in. Bonner sphere have been carried out in the 2.5 MeV neutron field at the accelerator facility of the Physikalisch-Technische Bundesanstalt in Braunschweig. The numerical and experimental results are presented. The measurements differ from the calculated values by a factor of about 1.09. It is questionable if these discrepancies can be explained by differences in the shielding, moderator and detector characteristics employed in the experiment and those assumed in the calculation. This is highlighting the importance of experimentally validating numerical MCNP response function calculations wherever possible.

Measurements of radionuclides in coal samples from two provinces of Pakistan and computation of External γ ray dose rate in coal mines

The radionuclides present in coal may not only be a health hazard for the coal miners but also may be a threat to the general population if these radionuclides disperse in the environment. This research has been conducted to quantify the radionuclides present in the coal samples from various coal-mines of two provinces, Punjab and Balochistan of Pakistan. In this regard, a high-purity Ge-detector-based γ-spectrometer was used. The maximum activity concentrations for 226Ra, 232Th and 40K were found to be 31·4±3·0, 32·7±3·2 and 21·4±5·0 Bq kg -1, respectively. A theoretical model to compute external γ-ray dose rate from a coal-mine surface was developed. The Monte Carlo simulation was employed to compute the required mass attenuation coefficients corresponding to the various γ-ray energies from 226Ra, 232Th, their progeny and 40K present in the coal samples. In addition, the effective thickness of coal slab for self-absorption was also computed using the Monte Carlo Neutron Photon (MCNP) transport code. The computed external γ-ray dose rate has been found to be much below the dose ratelimits for occupational persons as well as for the general population.

Experimental and simulation study of neutron dosimetry at various neutron energies

A Monte Carlo Neutron Photon (MCNP) transport code has been employed to simulate CR-39 plastic track detector as neutron dosimeter at various neutron energies. In each simulation a monoenergetic neutron source of a particular energy was embedded in the center of a sphere made of CR-39. Surrounding the source there were concentric shells of 2 μm CR-39 track detectors. The code, MCNP, was run on personal computer for 7.5×10 6 histories. The number of proton recoils in each shell of 2 μm of CR-39 were determined. The simulation results show that apart from proton recoils (for E n > 1.0 MeV), about 50% recoils are due to heavy charged particles i.e. Oxygen and Carbon in CR-39. This indicates that etched tracks are not only due to recoil of protons but also due to recoil of heavy charged particles. The upper limits of the track registration efficiencies have been determined as a function of neutron energies. These simulation results have been experimentally verified using CR-39 track detectors at various known energies of neutrons. The proton and heavy charged particle recoil tracks in CR-39 were made visible by etching in NaOH solution, at 70±1°C for 1.75 hours.

Treatment planning figures of merit in thermal and epithermal boron neutron capture therapy of brain tumours

The boron neutron capture therapy (BNCT) figures of merit of advantage depth, therapeutic depth, modified advantage depth and maximum therapeutic depth have been studied as functions of 10B tumour to blood ratios and absolute levels. These relationships were examined using the Monte Carlo neutron photon transport code, MCNP, with an ideal 18.4 cm diameter neutron beam incident laterally upon an ellipsoidal neutron photon brain-equivalent model. Monoenergetic beams of 0.025 eV (thermal) and 35 eV (epithermal) were simulated. Increasing the tumour to blood 10B ratio predictably increases all figures of merit. 10B concentration was also shown to have a strong bearing on the figures of merit when low levels were present in the system. This is the result of a non-10B dependent background dose. At higher levels however, the concentration of 10B has a diminishing influence. For boron sulphydryl (BSH), little advantage is gained by extending the blood 10B level beyond 30 ppm, whilst for D,L,-p-boronophenylalanine (BPA) this limit is 10 ppm. To achieve a therapeutic depth of 6 cm (brain mid-line from brain surface) using the thermal beam, a tumour to blood ratio of 25 with 10 ppm 10B in the blood is required for BPA. Similarly, a tumour to blood ratio of 8.5 with 30 ppm blood 10B is required for the maximum therapeutic depth of BSH to reach the brain midline. These requirements are 5 times above current values for these compounds in humans. Applying the epithermal beam under identical conditions, the therapeutic depth reaches the brain midline with a tumour to blood 10B ratio of only 5.7 for BPA. For BSH, the maximum therapeutic depth reaches the brain midline with a tumour to blood ratio of only 1.9 with 30 ppm in the blood. Human data for these compounds are very close to these requirements.

Analysis of Tritium Production in a Sphere of6LiD with Oralloy Core Irradiated by 14-MeV Neutrons

Tritium production in a sphere of [sup 6]LiD with an Oralloy core irradiated by a central source of 14-MeV neutrons has been calculated and compared with experimental measurements. The experimental assembly consisted of an Oralloy sphere surrounded by three solid [sup 6]LiD concentric shells with ampoules of [sup 6]LiH and [sup 7]LiH located in several positions throughout the assembly. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport through-out the system and tritium production in the ampoules. The MCNP calculations were three-dimensional and employed ENDF/B-V cross sections. The overall experimentally observed-to-calculated ratios of tritium production were 0.996 ([plus minus]2.5%) in [sup 6]LiH ampoules and 0.903 ([plus minus]5.2%) in [sup 7]LiH ampoules. Tritium production in a sphere of [sup 6]LiH without an Oralloy core has been reanalyzed using ENDF/B-V cross sections, and the results are reported. The reanalyzed observed-to-calculated values of tritium production were 1.053 ([plus minus]2.1%) in [sup 6]LiH and 0.999 ([plus minus]2.1%) in [sup 7]LiH. The foregoing several uncertainties do not include an estimated [lt]6% systematic error in the observed values.

Neutron and Gamma-Ray Spectra from a Variety of Materials Bombarded with 14-MeV Neutrons

Measurements of the neutron and gamma-ray leakage spectra from 15 spherical target assemblies (carbon, nitrogen, H2O, C2F4, aluminum, silicon, titanium, iron, copper, tantalum, tungsten, gold, lead, 232Th, and 238U) pulsed with 14-MeV neutrons were made using time-of-flight techniques. The spheres were ∼30 g/cm2 thick to maximize the gamma-ray leakage per central source neutron. Among all the materials studied, silicon shows the highest conversion factor (∼2 γMeV/n), and lead the lowest (0.31 γMeV/n). Monte Carlo neutron-photon transport calculations were done using the TART and SANDYL codes, with the ENDL and ENDF/B-V libraries. Comparisons with the neutron measurements confirm earlier results, where both libraries reproduced the leakage spectra for most of these materials reasonably well. The gamma spectra calculated with ENDL give a fair representation of the measurements, with the exception of the initial calculations for 16O(H2O) and 19F(C2F4), where serious discrepancies are found. Improvements were obtained for 16O after a re-evaluation of the neutron-induced cross sections based on more recent microscopic experimental data. This was also the case for 19F, where the calculations now overestimate the measurements by 30%. Calculations with the ENDF/B-V are lower than the experimental measurements for most of the materials.

Reanalysis of tritium production in a sphere of6lid irradiated by 14-MeV neutrons

Abstract Tritium production in a sphere of 6LiD irradiated by a central source of 14-MeV neutrons reported by Hemmendinger et al.1 has been reanalyzed. The 6LiD sphere sonsisted of 10 solid hemispherical shells nested with ampules of 6LiH and 7LiH located at 5.0, 7.7, 12.6, 20.0, and 30.0 cm from the center. The Los Alamos Monte Carlo Neutron Photon Transport Code was used to calculate neutron transport through the 6LiD and tritium production in the ampules. The model was three dimensional and employed ENDF/B-V cross sections.