MCNP Code Research Articles

Estimation of H*(10) and E(Ω) (and H*) using the Hp(10) and Hp(0.07) detectors of a body TLD

The H*(10) and E(Ω) responses to photons from 15 to 1253 keV were determined as a function of source orientation for the 7LiF:Mg,Cu,P ‘body’ and ‘skin’ elements of a thermoluminescence dosemeter optimized to measure Hp(10) and Hp(0.07), and are presented here relative to the results from 662 keV calibration exposures. The calculations were performed using the Monte Carlo transport code MCNP6.2, with the dosemeter located either free-in-vacuo or on a concrete wall; the impacts of the radiation field and dosemeter mounting are discussed. The relative response profiles were found to be fairly flat at energies above a few 10s of keV, leading to the conclusion that the personal dosemeter may be used in some environmental or workplace fields to provide reasonable estimates of ambient dose equivalent and effective dose; the latter is additionally relevant given the newly proposed operational quantity ambient dose (H*). Optimal correction factors were also derived that allow the dosemeter's standard Hp(d,0°,137Cs) system calibrations to be maintained.

A comparative Monte Carlo simulation study on shielding features of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass system against X-ray by GEANT4 and MCNPX codes

Regarding to their unique physical and mechanical features, glasses and glass–ceramics are suitable materials for shielding purposes. The present study evaluates the shielding properties of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass system using Monte Carlo GEANT4 and MCNPX codes for X-ray radiations with an energy range of 20 to 100 keV. MAC values of the Ta0, Ta1, Ta2, Ta2.5, and Ta3 samples of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass were computed using Phy-X/PSD, GEANT4, and MCNPX codes and compared. According to the results, the programs have good compatibility with each other. For instance, in the energy of 40 keV and for the Ta2 sample, GEANT4 and MCNP codes are 1.445765406 and 1.517801204 cm2/g, respectively, indicating 7.419529525 and 2.829628418% differences with 1.562 cm2/g obtained using the Phy-X/PSD software. According to recent estimations, the Ta3 sample of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass system can be selected as the best shield compared with the other samples.

The role of MoO3 on the physical, elasto-mechanical and nuclear shielding efficiency of barium-boro-bismuthate glass system: Comparative investigation

Investigations were conducted on the effect of molybdenum addition on the physical, elasto-mechanical and nuclear shielding characteristics of five different glasses, designated S1 through S5, with varying amounts of MoO3. The mechanical parameters of the examined glasses, including their Young (Y), bulk (B), shear (S) and longitudinal (L) moduli, microhardness (H), and poisons ratio, were calculated utilizing the Makishima-Mackenzie technique. The physical and mechanical characteristics of the examined glasses are improved by the addition of MoO3. The Y, B, S, and L moduli rose with increasing MoO3 concentrations from 34.6680 to 35.9989 GPa, 14.13578–14.664 GPa, 21.10714–22.60943 GPa, and 31.70897–33.59673 GPa, respectively. The H values of the glass network also dropped from 0.016431 to 0.016885 GPa. Moreover, as the concentration of MoO3 rose, the Poisson ratio value changed from 0.427021 to 0.429533. The MCNP5 code and WinXCOM program were used to compute various shielding parameters for X/gamma-rays in the energy range between 0.015 and 15 MeV. A good agreement was found between the estimated values using both MCNP5 code and WINXCOM program. The S5 glass sample exhibited the greatest gamma-ray attenuation characteristics among the examined glasses. When compared to other radiation shielding materials including recently studied glasses and polymers as well as other typical gamma ray shielding materials, the current glass composition offers complete (100 %) x-ray protection and effective efficacy for gamma-ray shielding at energies up to 15 MeV. Calculations were made of the studied glasses' macroscopic removal cross sections (∑R), which were then compared to those of commonly used neutron shielding materials. S1 sample performed the best ∑R value while S5 displays a value (0.10516 cm−1) that is comparable to those of graphite, concrete, basalt magnetite, and a few other recently studied glasses and alloys. The heavy ions (proton and alpha particle) attenuation parameters have been evaluated in energy range between 0.01 and 15 MeV. Regarding the studied glasses, the S5 sample was shown to have the best proton and alpha shielding properties. These results imply that the glass samples under study make good nuclear shielding materials for a range of applications.

Calibration and performance evaluation of an N-type HPGe detector for a collimated and focused PGNAA system by experimental and Monte Carlo methods

The MCNP code was utilized to assess the operational status of an N-type high-purity germanium (HPGe) detector and to evaluate its potential for a collimated and focused Prompt Gamma Neutron Activation Analysis (PGNAA) detection system. In order to calibrate the detector, a narrow beam γ collimator was designed to scan the detector in both axial and radial directions. The curvature of the crystal front edge and the non-uniform description of the detector dead layer (DL) were taken into account when adjusting the effective crystal size. A collimated 137Cs γ source was used to assess the detection efficiency in the scanning experiments. Monte Carlo studies were performed using 0.662 MeV γ rays, and separately, a uniform γ-ray energy distribution from 0 to 1.6 MeV. Lastly, the effects of a collimation aperture with different sizes and different incident directions were considered to discuss the relationship between the spatial resolution and suitable measurement scenarios. Therefore, a detailed calibration method for non-uniform dead layers was introduced through a comparison between experiments and Monte Carlo simulations. This work reevaluates the performance of HPGe detectors and provides valuable data support for a detailed system design and γ spectroscopy analysis in future.

Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dual-purpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF–ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.

Estimating spent fuel burnup with Neutron measurements: A Practical Rule of Thumb Equation

We present a concise equation correlating the burnup of spent nuclear fuel (SF) with the neutron count rate, developed through comprehensive data analysis from Origen-ARP and MCNP codes. This equation is applicable to Fork detectors, commonly used for verifying SF assemblies before their transfer to new storage sites. The detector is assumed to be a helium-3 detector. The reaction rate (RR) in the 3He detector is influenced by Total Neutron Source Intensity (TNSI), net neutron multiplication, and neutron capture during transit to the detector. TNSI emerges as the most influential factor. Two scenarios were explored: one involving pure water and the other with water containing 2000 ppm of boron. The characteristics of the concise equation are also analyzed.

Investigation of nano MgO loaded polyvinyl chloride polymer in protective clothing as a nonlead materials

Recently, investigation of advanced shielding materials to be used as an alternative to lead apron has become important. In the current study, MgO loaded into PVC matrix as a non-lead modern shielding composite was modeled to evaluate its performance on radiation protective clothing (RPC). Parameters such as mass attenuation coefficient (MAC), mean free path (MFP), flux buildup factor (FBF), transmission factor (TF) and lead equivalent value (LEV) of samples were calculated using MCNPX Code. The simulation of the MCNP code was validated, by comparing the mass attenuation of concrete sample, with standard XCOM data and very good agreement was attended between XCOM and MC Code results. The MAC of nano and micro-sized samples were also compared with pure PVC and it was found that the nano MgO particle exhibits higher attenuation compared to micro MgO particle and pure PVC. The results show that, the MAC of samples increased to 63.13 % in 1.332 MeV with increasing filler concentration of nano MgO to 50 wt% relative to pure PVC. Investigation of LEV shows that nano MgO sample has more effective than Pb in 1.173 and 1.332 MeV gamma ray energy so that it provides 36.46 % and 11.13 % lighter RPC than Pb ones.

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

Monte Carlo analysis of dosimetric issues in space exploration

The Radiation protection is of paramount importance in the planning of human exploration activities in space. The related risks must be considered with respect to two aspects: devising a proper shielding and providing answers to the requirement of an effective dosimetry evaluation in astronaut's activities. Both aspects have been considered using the Monte Carlo (MC) code MCNP 6.2 as the reference tool. As case study an application devised for the National Aeronautics and Space Administration (NASA) Artemis program has been chosen. The project aims to establish a sustainable human presence on the Moon, envisioning the realization of an outpost that will serve as a steppingstone for space exploration endeavors. A Class III shelter, in situ resource utilization (ISRU) built habitat for the Moon, has been designed through computational methods and topology optimization techniques, and analyzed in terms of radiation shielding performances and the strictly related structural behavior. The outpost must be able to withstand temperature variations, micrometeorite impacts, and the absence of a substantial atmosphere. Any solution studied to respect the constraints must devise robust and innovative materials and techniques to create habitats that have as goal the shielding from the Galactic Cosmic Rays (GCR) and from the solar flares to provide a safe and habitable environment at the time scales scheduled for the missions. Moreover, the outpost design must incorporate strategies for extracting and utilizing local resources. Overcoming such challenges will pave the way for the establishment of a sustainable human presence on the Moon and serve as a crucial leap for future space exploration missions.

The feasibility of K XRF bone lead measurements in mice assessed using 3D-printed phantoms

This article describes the development of a system for in vivo measurements of lead body burden in mice using 109Cd K x-ray fluorescence (XRF). This K XRF system could facilitate early-stage studies on interventions that ameliorate or reverse organ tissue damage from lead poisoning by reducing animal numbers through a cross-sectional study approach. A novel mouse phantom was developed based on a mouse atlas and 3D-printed using PLA plastic with plaster of Paris ’bone’ inserts. PLA plastic was found to be a good surrogate for soft tissue in XRF measurements and the phantoms were found to be good models of mice. As expected, lead detection limits varied with mouse size, mouse orientation, and mouse position with respect to the source and detector. The work suggests that detection limits of 10 to 20 μg Pb per g bone mineral may be possible for a 2 to 3 hour XRF measurement in a single animal, an adequate limit for some pre-clinical studies. The 109Cd K XRF mouse measurement system was also modeled using the Monte Carlo code MCNP. The combination of experiment and modeling found that contrary to expectation, accurate measurements of lead levels in mice required calibration using mouse-specific calibration standards due to the coherent scatter peak normalization failing when small animals are measured. MCNP modeling determined that this was because the coherent scatter signal from soft tissue, which until now has been assumed negligible, becomes significant when compared to the coherent scatter signal in bone in small animals. This may have implications for some human measurements. This work suggests that 109Cd K x-ray fluorescence measurements of lead body burden are precise enough to make the system feasible for small animals if appropriately calibrated. Further work to validate the technology’s measurement accuracy and performance in vivo will be required.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Study on neutronic behavior of VVER-1000 fuel assembly with duplex fuel rod design

Various studies suggested that duplex fuel rods have better neutronic characteristics than ordinary single-oxide fuel pellets. A duplex fuel rod consists of an inner uranium dioxide (UO2) layer and an outer thorium dioxide (ThO2) layer. This paper aims to study the neutronic behavior of the VVER-1000 fuel assembly with duplex fuel rod design. The duplex fuel rods were loaded into the fuel assembly in three configurations: outermost region, mid-region, and central region of the assembly. Three compositions with volume fractions of 75% UO2-25% ThO2, 70% UO2-30% ThO2, and 65% UO2-35% ThO2 were considered. A series of numerical calculations were conducted using the MCNP6 code and ENDF/B-VII.1 library. In general, there is an improvement in DTC and MTC value for fuel assembly using Duplex fuel rods. However, βeff and ℓ values tend to decrease while the value of Λ tends to increase which may affect fuel assembly response during transient scenarios. The results show that Configuration I can maintain its criticality for up to 15 MWd/kg for all fractions of UO2-ThO2 being used in duplex fuel, while Configuration II and Configuration III only reach below 15 MWd/kg when using duplex fuel rods. The Doppler temperature coefficient (DTC) and the moderator temperature coefficient (MTC) become stronger with the increasing ThO2 volume fraction within the VVER fuel assembly. Compared with non-duplex fuel, the amount of 239Pu was lower when using duplex fuel rods, but minor actinides was slightly higher at 40 MWd/kg. It can be concluded that the VVER-1000 with a duplex fuel rod could improve its safety characteristics besides that increasing fissile material content might be crucial to optimize VVER-1000 fuel management.

Neutronics analysis of the ion cyclotron resonance heating antenna of the China Fusion Engineering Test Reactor

Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.

Impurity sensitivity analysis for activated structure materials of OPR-1000

Various kinds of radioactive wastes are expected to be generated resulting from the decommissioning of nuclear power plants. For the safe management of decommissioning wastes, the preliminary classification of radioactive wastes such as very low level, low level, intermediate level, and regulatory clearance has to be performed. In this regard, we intend to conduct an activation analysis and radioactive waste level estimation using a computational simulation code for the structure materials of an OPR-1000, a standard nuclear power plant in South Korea. In addition, the structure material that constitutes the reactor vessel, reactor vessel internals, and bio-shield concrete of the nuclear power plant contains small amounts of unintended impurities. The results of the activation analysis may vary depending on the presence of those impurities. In this study, a detail modeling of nuclear reactor and the activation analysis are carried out combined the MCNP code with the ORIGEN-S code. Based on this methodology, the sensitivity test is done for impurities of structure materials. From the result of the specific activity of activated structure material, the cobalt impurity is found to be the most important contributor in the level determination of radioactive wastes. However, the bio-shield concrete has the highest specific activity when Li impurity is contained. From this study, some impurities affect radioactive waste level, thus it is recommended to consider structure material impurities for decommissioning the nuclear power plants.

Experimental and Monte Carlo Evaluation of Spherical Ionization Chambers’ Response to 137Cs Radioactive Source

Measurements were performed for the calibration of an ionization chamber in the secondary standard dosimetry laboratory of IFIN-HH. A 10-liter spherical ionization chamber was calibrated using a 1-literspherical reference standard ionization chamber. The ionization chamber’s response to 137Cs was also evaluated computationally by MCNP and FLUKA Monte Carlo codes. The models of the experimental setup are validated and can be used for further studies, especially in cases where the real experiment is hard or impossible to perform

Measurements of AmBe and PuC spectra using EJ309 and stilbene scintillators

Two organic scintillators (EJ309 and stilbene) with their different detection chains and elaborate software were used to measure neutron spectra of AmBe and PuC sources. Method of shadow cone ensures minimization of scattered neutron influence. To separate neutrons and γ, a modern method of pulse shape discrimination was employed using fast digitizers. Calibration was performed by γ sources and three different methods for the identification of the Compton edge. Spectrometric systems also used different light output functions and unfolding methods. The same spectra were calculated using the MCNP6 code and the JENDL-5 library. The measured and calculated results were mutually compared, and also with the published results available in the literature. Generally, the results agree on the main peaks of neutron spectra but differ in their abundance. The lower limit for the ability to distinguish between neutrons and γ in spectrometry with organic scintillators was found at about 1.5 MeV.

Design of proton-beam degrader for high-purity 89Zr production

This work investigated the most suitable type of degrader (Cu, Al or Nb) and its thickness, taking into consideration the salient aspects of concrete activation for high-purity 89Zr production by 89Y(p,n)89Zr nuclear reaction. The MCNP and FISPACT codes were used to determine the optimal degrader thickness and the radioactivity of shielding concrete by neutron activation, respectively. The results showed that the optimal thickness of the beam degraders was 1.16, 3.19, and 1.33 mm for Cu, Al, and Nb, respectively. The neutron production rate per proton and the energy and angular distributions of neutrons varied depending on the type of degrader. Considering the radioactivity of the target-room concrete and the amount of radioactive waste expected to be generated, the use of a 1.33-mm-thick Nb degrader for 89Zr production was determined to be the best choice.

Optical structural and radiation shielding efficiency of vanadium-doped LZBS glass

By using the classic melt-quench method, a glass system including silicon-doped vanadium ions, lithium, zinc, and boron was fabricated with the following formula, 27SiO2-2ZnF2-53B2O3- xV2O5-(18-x) Li2O- with x = 0.0, 0.25, 0.5, 1, and 2 mol%. XRD was used to confirm amorphous status. Density, molar volume, oxygen packed densities and Inter-nuclear distance were determined. The optical characteristics were investigated. The energy optical-gap declines with increased vanadate’s content. It is 3.25 eV for pristine glass and reduced to 1.91 eV for glass doped with 2 mol% of vanadium ions. Refractive index, absorption coefficient, dielectric constant, optical and electrically conductive, basicity and oxide polarizability have been estimated. On the other hand, the glass samples were simulated using MCNP-5 code to set their radiation shielding ability. The results of the coefficients of attenuation μm, μL, HVL, MFP, Zeff, Zeq, Ne, RPE and FNRCS were calculated. The comparison between results obtained by the MCNP5 code, PHYX-PSD package and XCOM at photons flux of 0.015 MeV to 15 MeV for all glass samples were in decent accord. All findings of the parameters refer to the good shielding abilities of our glass system when increasing the concentration of (V2O5) at the expense of (Li2O) values at the different energies.

Investigation of Tissue Components Impacts on Dose Enhancement Factor Using Monte Carlo Code

Despite the progress of science in cancer treatments and radiotherapy improvements, there are still several side effects that occur during tumors treatment, particularly on healthy tissues surrounded tumors. Newer treatment methods are being explored lately, one of which is the use of nanoparticles, wherein the tumor is injected with gold nanoparticles. Its aim is to enhance tumor sensitivity to radiation and reduce radiation damage to healthy tissues. Tissue type may play an effective role in enhancing the dose being received under the use of nanoparticles. This study aims to find the effect of different tissue components on dose enhancement factor through MCNP6 and GATE simulations, as well as to accurately compare the simulation results of these two code packages for dose enhancement factors. A 125I brachytherapy source was simulated in phantoms for five tissues or materials (adipose tissue, breast tissue, soft tissue, water, and brain tissue). MCNP6 simulation code was validated by comparing its results with a previous study by Cho et al. Gold nanoparticles were injected as a mixture at a concentration of 7 mg/g into tissues inside a tumor. MCNP6 and GATE simulation results were compared. It was estimated from MCNP simulations that the highest radiation dose enhancement of 2.34 occurs in adipose tissue while lowest dose enhancement of 1.69 is in brain. In comparison, from GATE results, the estimates were that the highest value of dose enhancement factor also occurred in adipose tissue at 2.01, and the lowest value in brain at 1.48. The comparison between two codes suggest that they are compatible with the percentage difference in all tissues being less than 15 %. This study confirms that both MCNP6 and GATE codes could calculate DEF for different tissues under irradiation from a low-energy source.

Estimating North Korea's nuclear capabilities: Insights from a study on tritium production in a 5MWe graphite-moderated reactor

This study explores the potential for tritium production in North Korea's 5MWe graphite-moderated reactor, a facility primarily associated with nuclear weapons material production. While existing research on these reactors has largely centered on plutonium, our focus shifts to tritium, a crucial element in boosted fission bombs. Utilizing the MCNP6 code for simulations, the results estimate that North Korea can possibly produce approximately 7–12 g of tritium annually. This translates to the potential production of 1–3 boosted fission bombs each year. By incorporating tritium production into assessments of North Korea's nuclear capabilities, our methodology provides insights into the dynamics of the country's nuclear force, revealing a more diversified and complex composition than previously assumed. The findings significantly aid policymakers, regulatory bodies, and researchers in comprehending potential proliferation risks associated with graphite-moderated reactors and in developing strategies to address the nuclear threat emanating from North Korea.