Monte Carlo Code MCNP Research Articles

Nuclear performances of the water-cooled lithium lead DEMO reactor: Neutronic analysis on a fully heterogeneous model

The development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO) is a key issue within the EUROfusion roadmap. The DEMO reactor is designed to produce a fusion power of about 2 GW and generate a substantial amount of electricity, relying on a closed tritium fuel cycle: it implies that the breeding blanket (BB) shall guarantee a suitable tritium production to enable a continuous operation without any external supply.The Water-Cooled Lithium Lead (WCLL) concept is a candidate for the DEMO BB: it uses liquid Lithium Lead as breeder and neutron multiplier and water in PWR condition as coolant. The neutronics analyses carried out in the past have been performed using a semi-heterogeneous representation of the BB, since the complexity of its structure makes the generation of a detailed MCNP model a very demanding and challenging task. Results highlighted good performances for the WCLL BB, both in terms of shielding effectiveness and tritium self-sufficiency.A recently updated assessment of the tritium breeding ratio (TBR) requirement for DEMO, considering margins for calculation uncertainties and incomplete models of the whole machine, led to the definition of a tentative 1.15 value for the TBR. Moreover, the implementation of an accurate BB neutronics model, consistent with the engineering design, is recommended for the evaluation of the tritium self-sufficiency. In order to tackle these issues, an MCNP model of the DEMO tokamak, integrating a fully heterogeneous WCLL BB has been developed for the first time, including an accurate description of the FW water channels, as well as a comprehensive definition of the breeding zone inner structure. A complete assessment of the WCLL BB nuclear performances, through 3D neutron and gamma transport simulations, has been carried out by means the MCNP Monte Carlo code and JEFF nuclear libraries.

Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

The BEAVRS benchmark provides detailed design data and in core flux measurements for a standard PWR. In this work the BEAVRS benchmark is simulated at hot zero power using the Monte Carlo code MCNP6. The effective multiplication factor is estimated at various control banks insertions and boron concentrations. Calculations also include control rod banks worth, and isothermal temperature coefficient. Axially integrated Thermal flux in 58 assemblies, resembling detector positions in the core, are also evaluated and compared to the actual data provided by the benchmark. Axial thermal flux calculated for selected assemblies was compared to the results produced by detector signals located at 61 axial positions. Radial power distribution in the whole core is also evaluated, and compared to a previous study. The accuracy of thermal flux and radial power calculations were evaluated using two methods; the absolute relative difference and the root mean square deviation. The model was capable of simulating the benchmark with good degree of accuracy.

Neutron shielding characteristics of polymer composites with boron carbide

In this study, the neutron shielding capabilities of boron carbide (B4C)-reinforced polyester-matrix composite materials were investigated. For the study, 20- 40- and 60-mm-thick specimens were first prepared by reinforcing with 10–50% B4C, and later their neutron shielding characteristics were studied by using various characterization techniques. The shielding capability of the composite materials was observed to increase with specimen thickness. In addition, increased amounts of B4C reinforcement improved the attenuation ratio of the material thereby increasing the shielding effectiveness. This trend, however, changed when the B4C ratio reached 40% and 50% at which a gradual decrease in neutron attenuation was observed. The experimental setup was also simulated using the Monte Carlo code MCNP-5, and similar computations were carried out. The experimental data almost completely agreed with computational results.

Application of the MTSVD method in the experimental x-ray spectrum unfolding in the diagnostic energy range using CR film and Plexiglas wedge

Obtaining the quality of the beam is considered one of the most important quality control issues for X-ray tubes. X-ray spectrum is a very important aspect of quality control of medical imaging devices. However it may not be measured directly due to the very high photon flux, increasing the problems of the dead time and the Pile-up effect. The purpose of this study is to use a simple, and indirect method to unfold the primary X-ray spectrum. In this investigation, a system composing of a PMMA wedge, and a CR film was designed and constructed. The absorbed dose by the film is related to the primary spectrum by response matrix which was obtained using the MCNP Monte Carlo code exactly. The response matrix was used to unfold the spectrum, from the reading of the CR film. However, as the response matrix is ill-conditioned, it is not possible to solve the equation system in the normal way. Therefore the MTSVD regularization method in MATLAB software was used for unfolding the primary spectrum. The shape of the spectrum and the beam quality was compared with the IPEM 78 report catalog. The results indicate that the X-ray spectrum may be unfolded using this indirect method with sufficient accuracy.

Application of deterministic sampling methods for uncertainty quantification in manufacturing tolerances in neutron physics

In this work, an application of the deterministic sampling (DS) methodology for uncertainty quantification of manufacturing tolerances is assessed. The DS methodology is applied to eight benchmark cases, from two criticality safety benchmark series, extracted from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (i.e., LCT-007 & LCT-039). The DS results are compared to two other sampling techniques: Random Sampling and Latin Hypercube Sampling. Ten input parameters such as isotopic composition of the fuel, geometry properties of the assembly, pitch array, etc. were considered. Only the multiplicative factor (keff) was considered as output quantity. Evaluation of the keff was obtained using the Monte Carlo code MCNP for the LCT-007 case, and the Serpent code for LCT-039.Results show very good agreement among all methodologies for the calculated keff values and their corresponding uncertainties. The largest uncertainty difference with DS, compared to the reference values, is 46 pcm for LCT-007 and 9 pcm for LCT-039. The use of DS remarkably reduces the number of samples needed to perform an accurate uncertainty quantification (from thousands of samples down to a dozen).This study provides insights in the application of DS methodology in the neutron physics domain and showcases its benefits and limitations.

The effect of incorporated recycled low density polyethylene (LDPE) on the fast neutron shielding behaviour (FNSB) of clay matrix using MCNP and PHITS Monte Carlo codes

In order to develop a cheaper, environmentally friendly and self-sustaining high hydrogen content material for fast neutron shielding, clay-polyethylene composites containing different weight fractions of recycled low density polyethylene (LDPE) were fabricated. The effect of the incorporated LDPE on the fast neutron shielding behaviour (FNSB) of the clay matrix was investigated using MCNP and PHITS Monte Carlo codes. This was done by evaluating the fast neutron effective removal cross section (ƩR), mean free path (λ) and the neutron yield of the fabricated composites. The obtained values of effective removal cross section ƩR for the composites range between 0.056 cm−1 and 0.079 cm−1, while the values of mean free path λ are between 17.857 cm and 12.658 cm. The results show that the effective removal cross section of the samples significantly increased, while the mean free path decreased as the weight fraction of recycled LDPE increased. This implies that the fast neutron shielding behaviour of the composites improves as the LDPE concentration increases in the clay matrix. The neutron yield of the composites obtained PHITS code follows similar trend.

Developing A Method for Inter-Seed Effect Correction in 125I Interstitial Brachytherapy Using Artificial Neural Network

Introduction: Treatment planning systems use TG-43 dose calculation protocol for brachytherapy sources. Dose calculations based on TG-43 formalism do not correct the perturbations due to the presence of tissue inhomogeneity, applicators, and inter-seed effects. Inter-seed attenuation has an important effect on dosimetry in permanent implant brachytherapy. The aim of this study is to evaluate the inter-seed attenuation effect for I-125 permanent implants. Then, a software was developed to find the real dose distribution for different combinations of sources. Materials and methods: In the first step, a hypothetical generic source model was designed based on the configurations of different commercial source types. MCNP5 Monte Carlo code was utilized to simulate the single active generic source at the center of the phantom, and an inactive placed at various positions inside the phantom. An algorithm was introduced using artificial neural network models that can estimate the dose distribution in presence of inactive sources. Results: The Monte Carlo calculation results showed that the dose distribution is clearly affected by the inter-seed attenuation effect. Comparison of the artificial neural network results with the Monte Carlo simulation results show that the artificial neural networks can predict the inter-seed attenuation with acceptable accuracy, and with not much greater than 3% deviation.Discussion and conclusions: Inter-seed effect is dependent on the distance between the seeds. Decreasing distances would cause more effect. According to the results, it seems that the artificial neural network can be used as a tool for correction of inter-seed attenuation effect in treatment planning systems.

BENCHMARKING OF THE SERPENT 2 MONTE CARLO CODE FOR FUSION NEUTRONICS APPLICATIONS

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

METHODOLOGY FOR HIGH-FIDELITY DETERMINISTIC MODELLING OF SWISS LWR FUEL ASSEMBLIES

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated against a database of experimental information. Moreover, PSI has improved the pin power calculations by developing models of Swiss fuel assemblies for the Monte Carlo code MCNP, with the isotopic compositions obtained from the In-Core Fuel Management data of the Studsvik suite of codes, by using the SNF code. A step forward is to use a neutron code based on fast deterministic neutron transport methods. The method used in this work is based on a planar Method of Characteristics in which the axial coupling is solved by 1D SP3 method. The neutron code used is nTRACER. Thus, the methodology of this work develops nTRACER models of Swiss PWR fuel assemblies, in which the fuel of each pin and axial level is modelled with the isotopic composition obtained from SNF. This methodology was applied to 2D and 3D calculations of a Swiss PWR fuel assembly. However, this method has two main limitations. First, the cross sections libraries of nTRACER lack some of the isotopes obtained by SNF. Fortunately, this work proves that the missing isotopes do not have a strong effect on keff and the power distribution. Second, the 3D models require high computational memory resources, that is, more than 260 Gb. Thus, the nTRACER code was modified, so now it uses only 8 Gb, without any loss of accuracy. Finally, the keff and power results are compared with Monte Carlo calculations obtained by Serpent.

The buildup factor calculations of concrete with different proportions of CRT based on a BP neural network by MCNP

ABSTRACT Cathode ray tubes (CRTs) are glass materials that contain harmful heavy metal elements such as lead . CRTs have been widely used in electronic products sets for decades. At present, these discarded CRTs endanger the environment. Because the high-density heavy metal elements in CRTs have a good shielding effect for X/γ rays, these raw materials have become a good choice to produce radiation-proof concrete. In the selection and use of shielding materials, the buildup factor must be considered. Monte Carlo code MCNP is used to calculate the exposure buildup factor (EBF) of concrete with different proportions of CRT fragments in the photon energy range of 0.015–15 MeV for shielding thicknesses of up to 20 mean free paths (mfp). Back-propagation (BP) neural network is proposed to predict the EBF, and the prediction effect is evaluated, reliability of this method is verified; the average error is 3.4%, and the maximum error is 9.1%. Using this method, the EBF of concrete with an arbitrary proportion of CRTs, any energy level and any shielding thickness can be quickly obtained in the ranges of the parameters of the neural network training set (0.015–15 MeV and 0–20 mfp in this paper).

DNA damage in lens epithelial cells exposed to occupationally-relevant X-ray doses and role in cataract formation

The current framework of radiological protection of occupational exposed medical workers reduced the eye-lens equivalent dose limit from 150 to 20 mSv per year requiring an accurate dosimetric evaluation and an increase understanding of radiation induced effects on Lens cells considering the typical scenario of occupational exposed medical operators. Indeed, it is widely accepted that genomic damage of Lens epithelial cells (LEC) is a key mechanism of cataractogenesis. However, the relationship between apoptosis and cataractogenesis is still controversial. In this study biological and physical data are combined to improve the understanding of radiation induced effects on LEC. To characterize the occupational exposure of medical workers during angiographic procedures an INNOVA 4100 (General Electric Healthcare) equipment was used (scenario A). Additional experiments were conducted using a research tube (scenario B). For both scenarios, the frequencies of binucleated cells, micronuclei, p21-positive cells were assessed with different doses and dose rates. A Monte-Carlo study was conducted using a model for the photon generation with the X-ray tubes and with the Petri dishes considering the two different scenarios (A and B) to reproduce the experimental conditions and validate the irradiation setups to the cells. The simulation results have been tallied using the Monte Carlo code MCNP6. The spectral characteristics of the different X-ray beams have been estimated. All irradiated samples showed frequencies of micronuclei and p21-positive cells higher than the unirradiated controls. Differences in frequencies increased with the delivered dose measured with Gafchromic films XR-RV3. The spectrum incident on eye lens and Petri, as estimated with MCNP6, was in good agreement in the scenario A (confirming the experimental setup), while the mean energy spectrum was higher in the scenario B. Nevertheless, the response of LEC seemed mainly related to the measured absorbed dose. No effects on viability were detected. Our results support the hypothesis that apoptosis is not responsible for cataract induced by low doses of X-ray (i.e. 25 mGy) while the induction of transient p21 may interfere with the disassembly of the nuclear envelop in differentiating LEC, leading to cataract formation. Further studies are needed to better clarify the relationship we suggested between DNA damage, transient p21 induction and the inability of LEC enucleation.

Characterisation of neutron field in the polyethylene neutron irradiator

The paper describes neutron field characterisation of low-flux multipurpose educational irradiator developed at Czech Technical University in Prague. The irradiator is aimed for demonstration experiments including neutron activation analysis, delayed neutron counting, or studies related to neutron and/or gamma detection and spectrometric devices. It may accommodate various neutron sources including 252Cf or AmBe radionuclide sources, D-D generator, or in-reactor irradiated nuclear material serving as delayed neutron source, or gamma sources including radionuclide ones or short-lived sources produced in the adjacent reactor.The characterisation was performed based on neutron activation technique using gold foils. It included two experimental parts. The first one verified the level of response symmetry in the four irradiation channels and characterisation of the axial distribution in the irradiation channels. Within the second one, the responses to thermal and epithermal neutrons and the cadmium ratio in the central irradiation position were determined. Whereas the former was determined solely with the 252Cf source, the latter was performed for all available sources: 252Cf, AmBe, and the D-D generator. The experimental results were further compared to calculations by the MCNP Monte Carlo code.

Analysis of integrated target in thorium‐based molten salt fast energy amplifier

Conventional neutron targets in accelerator driven subcritical system (ADS) are based on the design using heavy metal like lead and lead-bismuth eutectic (LBE), which generally needs a separated loop for target cooling. In the thorium-based molten salt fast energy amplifier, an integrated target is conceived that the proton beam is directly placed in the chloride/fluoride molten salt core without a separated cooling loop for the target in consideration of the neutron productivity of the liquid molten salt fuel bombarded with the proton. The physical analysis of the spallation target is presented, including the spallation neutron yield, neutron spectrum, spallation products, and energy deposition. Chloride molten salt is a reasonable choice among four kinds of simulated target based on considering about the neutron yield, neutron spectrum, spallation products and energy deposition. The simulation code is the Monte Carlo code MCNP6.1. Besides, evaluation of the target size is presented, and two kinds of the beam tube layout are compared by considering the neutron economy in the system.

Measurement of delayed neutron emission from water activated by 14 MeV neutrons in a FW mock-up of ITER

Water activated by 14 MeV neutrons emits both high energy gammas from 16N and delayed neutrons from 17N. The isotopes 16,17N are produced through the 16O(n,p)16N (T1/2 = 7.13 s) and 17O(n,p)17N (T1/2 = 4.1 s) reactions, respectively. These sources of radiation represent one of the main issues for ITER having large impact on the design of the machine and on the qualification of components. To validate the calculation methodology used for ITER and to reduce the safety factors applied to the calculation results a water activation experiment was commissioned by F4E to ENEA and performed at the 14 MeV Frascati neutron generator (FNG). The experimental set-up consisted of a closed water loop, about 40 m long, where the cooling water, while transiting through an ITER FW mock-up, was irradiated by the D-T neutrons produced at FNG. 6.73 MeV gamma-rays from 16N and delayed (fast) neutrons from 17N were measured using a large CsI gamma-ray detector and an array of absolutely calibrated 3He detectors, respectively. This paper reports on the measurements of the delayed neutrons emission from 17N. The calibration of the 3He detector array and the experimental set-up are first discussed. Then the experimental results (E) for the delayed neutrons as well as the comparison with the calculation (C) performed using the MCNP Monte Carlo code (coupled to the FENDL3.1 data library) and the EASY-2007 inventory code are presented. The C/E ratio results close to unity, however it is affected by a very large error (> ±60 %) to be ascribed mainly to the uncertainty on the nuclear data used for calculation.

Assessment of DEMO WCLL breeding blanket primary heat transfer system isolation valve absorbed doses due to activated water

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the investigation of the absorbed dose induced by the decay of nitrogen radioisotopes produced by water activation, in the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits.The aim of this work is to assess the spatial distribution of the absorbed dose in the DEMO Upper Pipe Chase (UPC), focusing the attention on the space neighbouring a typical isolation valve of the Primary Heat Transfer System (PHTS), taking into account a period of 7 full power years. To this end, a computational approach has been followed adopting MCNP5 Monte Carlo code. In particular, a totally heterogeneous neutronic model of a portion of the UPC has been set up, including the valve and the main FW and BZ PHTS piping, and the spatial distribution of nitrogen isotopes concentrations, previously assessed, have been used to model the photonic and neutronic sources.The results obtained, herewith presented and critically discussed, provided some information on the nuclear issues of the WCLL BB PHTS, to be considered as hints for the blanket design optimization.

Developing a software for self-absorption correction for gamma spectrometry of soil and water samples based on MCNP5 Monte Carlo simulations

Gamma spectroscopy using HPGe is one of the most effective methods in determining the concentration of gamma emitting radionuclides in environmental samples. The purpose of this study is obtaining the efficiency calibration curves for the HPGe detector using MCNP5 Monte Carlo code, and designing appropriate software for correction of self-absorption caused by changes in density, height, and geometry of different samples. For this purpose the detector was simulated using MCNP5 Monte Carlo code, and the detector calibration curves were obtained for different geometries and heights, and appropriate software was designed for efficiency calibration. The results obtained in this study, show that changing the height, geometry, and density of the samples have significant effects on the detector efficiency because of the changes in self-absorption of the samples. Comparison of the self-absorption correction using the software, and the results of simulations show that designed software can predict the calibration curves for the new samples in different energies with error much less than 1%.

Investigation of the VVER-1000 reactor pressure vessel neutron fluence and displacement per atom using MCNP6

Ensuring the integrity of the reactor pressure vessel (RPV) during the operating life of nuclear reactors is of paramount importance, and hence the harmful effects of irradiation on the steels of the RPVs have been extensively studied especially the irradiation embrittlement effects. However, improvements in the calculation methods are still necessary to further reduce the discrepancy between calculations and experiments. This study was performed to examine the neutron fluence and respective rate of displacement per atom (DPA), an important parameter describing the irradiation-induced damage, in the VVER-1000 RPV steel based on the criticality calculation, instead of the common fixed source approach, with the MCNP6 Monte Carlo code, and identify the most sensitive locations in the VVER-1000 RPV. The MCNP6 modeling of the VVER-1000 reactor was verified against the OpenMC calculations in addition to the confirmation of the convergence of the effective neutron multiplication factor and the Shannon entropy of the fission source distribution in the MCNP6 criticality calculations to ensure reliable tally results of the neutron fluence. Consequently, the maximum neutron fluence in the VVER-1000 RPV was found at the positions adjacent to the peripheral fuel assemblies, and the respective DPA rate was analyzed in correlation with the neutron fluence and energy spectrum.

Scattering kernels for fast neutron therapy treatment planning

The University of Washington (UW) Clinical Neutron Therapy System (CNTS) has been used to treat over 3300 patients. Treatment planning for these patients is currently performed using an MV x-ray model in Pinnacle® adapted to fit measurements of fast neutron output factors, wedge factors, depth-dose and lateral profiles. While this model has provided an adequate representation of the CNTS for 3D conformal treatment planning, later versions of Pinnacle did not allow for isocentric gantry rotation machines with a source-to-axis distance of 150 cm. This restriction limited the neutron model to version 9.0 while the photon and electron treatment planning at the UW had moved on to newer versions. Also, intensity modulated neutron therapy (IMNT) is underdevelopment at the UW, and the Pinnacle neutron model developed cannot be used for inverse treatment planning. We have used the MCNP6 Monte Carlo code system to develop Collapsed Cone (CC) and Singular Value Decomposition (SVD) neutron scattering kernels suitable for integration into the RayStation treatment planning system. Kernels were generated for monoenergetic neutrons with energies ranging from 1 keV to 51 MeV, i.e. the energy range most relevant to the CNTS. Percent depth dose (PDD) profiles computed in RayStation for the CC and SVD kernels are in excellent agreement with each other for depths beyond the beam’s dose build-up region (depths greater than about 1.7 cm) for open 2.8 × 2.8 cm2, 10.3 × 10.3 cm2, and 28.8 × 32.8 cm2 fields. Lateral profiles at several depths, as well as the PDD, calculated using the CC kernels in RayStation for the full CNTS energy spectrum pass a 3%/3 mm gamma test for field sizes of 2.8 × 2.8 cm2, 10.0 × 10.3 cm2, and 28.8 × 32.8 cm2.

Simulations of scattered neutrons for the time-of-flight enhanced diagnostics (TOFED) neutron spectrometer on EAST

The time-of-flight enhanced diagnostics (TOFED) neutron spectrometer has been installed in TOFED house, which is a newly-built neutron diagnostic lab outside the EAST experimental hall. The line of sight (LOS) of TOFED is defined by the collimator going through the 150 cm biological shielding wall, which can reduce background neutrons very effectively for the neutron spectral measurements. The Monte Carlo code MCNP5 is used in the simulations to characterize collimation effects and the radiation safety of the new house. The collimator makes neutron flux almost fully perpendicular to S1 detectors. In addition, the proportion of scattered neutrons in the LOS of the TOFED is obtained for the measured spectral data interpretation. The simulation results show that the TOFED house satisfies the requirement for the TOFED operations at EAST and gives a safe environment for personnel access during the EAST operation.

A Stylized 3-D Benchmark Problem Set Based on the Pin-Fueled SmAHTR

In this paper, a set of stylized numerical benchmark problems is developed. These problems are based on the Oak Ridge National Laboratory preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, that uses prismatic fuel assemblies with cylindrical pins/rods containing tri-isotropic fuel particles. A detailed description of the benchmark problems is achieved by closing several outstanding design gaps and modifying the coolant channel shape to reduce bypass flow for improved coolant and fuel temperature distributions. The benchmark problems, while stylized, retain the important thermal-hydraulic and reactor physics features (e.g., fuel particles) necessary for benchmarking tools for reactor core analysis. In addition to the full description, detailed reference results such as the eigenvalue (keff) and fuel pin and assembly-averaged fission density distributions are provided for five benchmark problems: full-length fuel assemblies with control rods fully withdrawn and inserted, and full core with all control rods withdrawn, all control rods fully inserted, and some control rods fully inserted (near-critical core). The provided results are calculated using the continuous-energy Monte Carlo code MCNP.