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The nuclear radiation shielding materials are having a great importance in reactor shielding, nuclear laboratories, radiation sources storage, as well as in space radiation shielding. The experimental methods are time consuming, facility dependent, as well as costly affair in order to manufacture the shielding materials. The alternative approach is to use nuclear transport codes such as GEANT, MCNP, PHITS, FLUKA for nuclear radiation shielding application. The present work is planned to study the prediction capabilities of transport code for calculation of shielding parameters using two different radiation transport codes and comparison with experimental results. The simulation measurements were carried out by Monte Carlo analysis, 241Am–Be neutron source & neutron dosimeter, and High Density Polyethylene (HDPE) were modeled using MCNP 6 and PHITS 3.22 codes. Subsequently, experimental neutron dose rates were measured after different thickness of HDPE samples using live 241Am–Be neutron source and NRF31 dosimeter. The study shows a good agreement between the experimental and simulated results which suggest that the simulation method can be used for the optimization of new shielding composites. [ABSTRACT FROM AUTHOR]

In this study, a neutron and γ-ray absorbing (Gd 2 O 3 +W)/Al composite was fabricated by vacuum sintering. Firstly, through Monte Carlo Neutron and Photo Transport Code (MCNP) calculations, the content of Gd 2 O 3 and W in the composite was determined to be 10 wt% and 25 wt%, respectively. Then, the influence of sintering temperature and sintering time on the microstructure, density, compression properties and electrochemical corrosion behavior of the (Gd 2 O 3 +W)/Al composites were investigated. The results indicate that sintering temperatures below 600 °C and sintering times shorter than 2 h led to insufficient density of the composite. Conversely, sintering temperatures above 600 °C and sintering times longer than 2 h caused W to agglomerate, resulting in a decrease in the density of the composite and a deterioration of its electrochemical performance. Therefore, the optimal sintering temperature and sintering time for the (Gd 2 O 3 +W)/Al composites were determined to be 600 °C and 2 h, respectively. Finally, the immersion corrosion performance in H 3 BO 3 solution of the composite fabricated by the optimal process parameters was studied in detail. It was found that pitting was the main corrosion form of the (Gd 2 O 3 +W)/Al composite, and an oxidation film mainly composed of Al 2 O 3 was formed on the surface. [ABSTRACT FROM AUTHOR]

This study investigates the individual and combined effects of enhancing the radiation shielding properties of waste concrete using the optimal mix design of two waste material powders of different compositions. Marble (MD) and granite (GD) waste dust were individually utilized as partial replacements for cement at a replacement ratio of 6%. Furthermore, two additional mixes were prepared by incorporating 1% by cement weight of nano alumina (NA) to enhance the microstructure of the studied waste concrete. The MGA-concrete was analyzed using X-ray Fluorescence, Energy dispersive X-ray, X-ray diffraction analysis, transmission electron microscopy, and scanning electron microscope techniques. The radiation shielding assets of the examined Concrete samples, such as the linear attenuation coefficient (μ), half value layer (H1/2), tenth value layer (T1/10), and fast neutron removal cross-section were evaluated using the MCS5 Monte Carlo simulation algorithm and Phy-X software. The results showed that the linear attenuation for the GMN-concretes' order is CO < MD < GD < NA < MD + NA < GD + NA. The GD + Na concrete sample presents the best neutron performance. The studied GMN-concrete samples provide the best protection against γ-rays and fast neutrons. Lastly, the excellent performance of the mixes of waste Granite, Marble, and Nano-Alumina on ordinary would pave the way for their employment as radiation shielding in various nuclear and medical facilities. [ABSTRACT FROM AUTHOR]

AbstractMonte Carlo (MC) transport codes are a cornerstone of nuclear reactor analysis frameworks, providing reference solutions and multigroup cross sections, and even as core components in multiphysics couplings. These applications can be seen in toolkits from the Nuclear Energy Advanced Modeling and Simulation program and the U.S. Nuclear Regulatory Commission’s BlueCRAB (Comprehensive Reactor Analysis Bundle).Contrary to their ubiquitousness in reactor physics modeling and simulation, popular MC codes, such as MCNP, Serpent, and KENO, are still reliant on antiquated textual input formats. These input languages use a plethora of keywords with terse syntax to specify all facets of a model, including its geometry, materials, physics settings, and tally options. This poses a steep learning curve and a poor user experience. Being tied to unique text-based input formats also significantly complicates programmatic input generation or modification that may be desired and/or required within a multiphysics framework.This work demonstrates how the development of programmatic interfaces for MC codes can support model unification and translation activities. Building on the Python application program interface (API) development of MCNPy, similar capabilities are being implemented for Serpent that are able to support model translation. In future work, the Serpent API capabilities will be made more robust and the work will be further expanded to include translations with KENO. [ABSTRACT FROM AUTHOR]

Modeling the response of gamma detectors has long been a challenge within the nuclear community. Significant research has been conducted to digitally replicate instruments that can cost over USD 100,000 and are difficult to operate outside of a laboratory setting. The large cost and availability prevent some from making use of such equipment. Subsequently, there have been multiple attempts to create cost-effective codes that replicate the response of sodium-iodide and high-purity germanium detectors for data derivation related to gamma-ray interaction with matter. While robust programs do exist, they are often subject to export controls and/or they are not intuitive to use. Through the use of hybrid Monte Carlo methods, MATLAB can be used to produce a fast first-order response of various gamma-ray detectors. The combination of a graphical user interface with a numerical-based script allows for open-source and intuitive code. When benchmarked with experimental data from Co-60, Cs-137, and Na-22, the code can numerically calculate a response comparable to experimental and industry-standard response codes. Evidence supports both savings in computational requirements and the inclusion of an intuitive user experience that does not heavily compromise data when compared to other standard codes, such as MCNP and GADRAS, or experimental results. When the application is installed on a Dell Intel i7 computer with 16 cores, the average time to simulate the benchmarked isotopes is 0.26 s. Installation on an HP Intel i7 four-core machine runs the same isotopes in 1.63 s. The results indicate that simple gamma detectors can be modeled in an open-source format. The anticipation for the MATLAB application is to be a tool that can be easily accessible and provide datasets for use in an academic setting requiring gamma-ray detectors. Ultimately, this article provides evidence that hybrid Monte Carlo codes in an open-source format can benefit the nuclear community in both computational time and up-front cost for access. [ABSTRACT FROM AUTHOR]

As the availability and importance of high energy X-ray sources grows, accurate source characterizations provide critical information for field flatness corrections, beam hardening corrections, detector response corrections, and radiation shielding assessments. This study uses MCNP6.2 to create accurate high definition angular and energy dependent X-ray source definitions for the most common high energy industrial X-ray sources at 450 kVp, 3 MVp, 6 MVp, 9 MVp, and 15 MVp. [ABSTRACT FROM AUTHOR]

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.

Abstract This study investigates the individual and combined effects of enhancing the radiation shielding properties of waste concrete using the optimal mix design of two waste material powders of different compositions. Marble (MD) and granite (GD) waste dust were individually utilized as partial replacements for cement at a replacement ratio of 6%. Furthermore, two additional mixes were prepared by incorporating 1% by cement weight of nano alumina (NA) to enhance the microstructure of the studied waste concrete. The MGA-concrete was analyzed using X-ray Fluorescence, Energy dispersive X-ray, X-ray diffraction analysis, transmission electron microscopy, and scanning electron microscope techniques. The radiation shielding assets of the examined Concrete samples, such as the linear attenuation coefficient (μ), half value layer (H1/2), tenth value layer (T1/10), and fast neutron removal cross-section were evaluated using the MCS5 Monte Carlo simulation algorithm and Phy-X software. The results showed that the linear attenuation for the GMN-concretes’ order is CO

Titanium alloys play a vital role in optimizing the effectiveness and security of nuclear reactors, strengthening structural durability, and facilitating the effective handling of nuclear waste. The aim of this study is to investigate the gamma-ray, neutron, and transmission properties of four common titanium alloys through the examination of the deposited energy amount in the liquid sodium coolant material, in relation to the mechanical properties of these alloys. MCNP (version 6.3) is utilized for designing the titanium pipes. Next, the pipes were re-designed considering the elemental mass fractions and densities of the investigated titanium alloys. Grade 26 sample is reported with the highest values of mass attenuation coefficients and the lowest HVL values among those investigated alloys. Grade 26 is reported to have the lowest TF value, whereas Grade 12 demonstrated the highest TF value. The highest Effective Removal Cross Section (ΣR, 1/cm) value against fast neutrons is reported for Grade 26. The utilization of Grade 26 sample as pipe material resulted in the lowest deposited energy amount (MeV/g) and subsequent lowest contamination in the coolant material. Out of the alloys that were chosen for analysis, it has been determined that Grade 26 exhibits the highest level of strength. It can be concluded that the Grade 26 alloy exhibits desirable characteristics for applications in nuclear technologies that require superior gamma-ray and neutron absorption properties, as well as exceptional mechanical properties. Nevertheless, it is essential to emphasize the importance for ongoing studies to enhance the existing material properties of Grade 26, with the aim of achieving improved safety and efficacy in nuclear applications.

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR. [ABSTRACT FROM AUTHOR]

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.

Due to the radiation released by commonly used isotopes, many nuclear, medical, and industrial facilities require proper radiation shielding. In this work, distinct copper borate glasses intercalated with varied aluminum and vanadium oxide (Al2O3 and V2O5) content have been synthesized and used against radiation (gamma rays and fast/thermal neutrons). The different percents were as follows: [60% B2O3 + 35% CuO + (5-x)% Al2O3 + xV2O5], where x = 0, 1, and 2.5 wt.%, which was coded as BCu(5-x)Al:xV. The synthesized glass samples were characterized using Fourier transforms, infrared, and X-Raydiffraction analysis. Experimentally, the radiation shielding possessions of the samples were established using an HPGe detector at the gamma energy lines 0.356 MeV, 0.661 MeV, 1.173 MeV, and 1.332 MeV. Also, the prepared glasses were investigated theoretically using the Monte Carlo code (MCNP5) at photon energies of 0.015–15 MeV. Also, the fast and thermal neutron macroscopic effective removal cross-sections were calculated using MRCsC and JANIS-4.1 software, respectively. The prepared sample BCu2.5Al:2.5V, which has a vanadium and aluminum content of 2.5%, has the highest linear attenuation coefficient as well as the highest removal cross-section for fast, and thermal neutrons.

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.

AbstractThe Second Target Station (STS) at the Spallation Neutron Source of the U.S. Department of Energy’s Oak Ridge National Laboratory is being designed to become the world’s highest peak brightness source of cold neutrons. As the STS project evolves, neutronics and other engineering analyses will inform many design iterations. To facilitate this process, a fully automated optimization workflow was developed to convert a parameterized computer-aided-design model of the target into an unstructured mesh geometry model and then to run a neutronics calculation and (optionally) a mechanical analysis for each design iteration. This workflow enables efficient, high-fidelity modeling; simulation; and optimization of new designs, as has been demonstrated for the optimization of the STS neutron moderators. In this paper, we present the results of our first major effort to automate the design optimization process for a spallation target. In the first analysis, the goal is to find optimal dimensions of a monolithic tungsten target coupled with an optimal super-Gaussian proton beam profile to deliver maximum brightness of the resulting neutron beams while maintaining good mechanical properties of the target. In the second analysis, geometric and beam parameters are optimized for an alternative design with tungsten plates, which can reach superior mechanical performance without compromising the neutronics performance. [ABSTRACT FROM AUTHOR]

This research examines the effects of varying titanium oxide (TiO 2) concentrations in glass formulations, denoted as [(53-x)B 2 O 3 –7MgO–20BaF 2 –20Na 2 O-xTiO 2 ] or BMBNTix, where x varies from 0 to 8 mol% in 2 % increments, on optical properties and radiation shielding effectiveness. Using MCNP5 (MC) Monte Carlo simulations and Phy-X software, this study assesses a range of photon energies from 0.015 to 15 MeV, focusing on key radiation shielding metrics such as the linear attenuation coefficient (μ), the half-value layer (HVL), and the tenth value layer (TVL). The results reveal a clear relationship between increased TiO 2 levels in the BMBNTix glasses and improved radiation shielding performance. Notably, the linear attenuation coefficient (μ) steadily climbs from BMBNTi0 to BMBNTi8, highlighting the beneficial effect of a TiO 2 addition to the material's ability to attenuate radiation. This trend is inversely mirrored in HVL values, which decrease from 0.010 cm for BMBNTi0 to 7.646 cm for BMBNTi8, reflecting enhanced shielding efficiency over the entire studied photon energy range. The effective atomic number (Z eff) also showcases a dynamic range from 40.793 to 15.101, suggesting a complex relationship between TiO 2 content and photon interaction. Moreover, the study indicates a slight but consistent increase in the fast neutron removal cross-section (Σ R) from 0.104 cm−1 for BMBNTi0 to 0.106 cm−1 for BMBNTi8, pointing to improved neutron shielding with greater Ti and oxygen content in the glass. These insights highlight the potential of TiO 2 -enriched magnesium borate glasses as superior options for effective photon and neutron shielding, particularly in nuclear medicine, emphasizing their role in enhancing radiation safety measures. [ABSTRACT FROM AUTHOR]

In the work, we studied the influences of dead layer thickness on the whole gamma spectrum response for HPGe detector by using Monte Carlo simulation. From the simulated relation of the FEPEs and the dead layer thicknesses, the dead layer thickness of 0.567 mm was derived which increased compared with the nominal value of 0.46 mm from manufacturer after 4 years of operation. The multiple scattering region and peak decreased with different rates and the valley region of the forward-scattering photon increased quickly when the dead layer thickness increased. The energy dependence of the P/T and the general range was also carried out. The results showed that the effect of the increase in dead layer thickness on regions in the spectrum was not the same. Through this study, among several scintillators, it was possible to find a HPGe detector suitable for the purpose of the study. [ABSTRACT FROM AUTHOR]

AbstractIn the context of designing radiotherapy facilities, typical dose estimation methods involve analytical approaches, as outlined in International Atomic Energy Agency (IAEA) Safety Reports Series No. 47 (IAEA 47). These methods are known for their ease of use and rapid calculations, but they could lead to either overestimation or underestimation of radiation doses. Hence, the integration of Monte Carlo (MC) methods is considered valuable. In this particular study, a radiotherapy facility was modeled using MCNP version 6.2, and dose calculations were conducted using analytical techniques following both IAEA 47 guidelines and MC simulations. The study focused on monoenergetic photon cone beams with energies of 10 and 15 MeV. Notably, the beam’s orientation prevented primary radiation from reaching the dose location at the entrance of the maze, allowing only scatter radiation to contribute to the tally. Given the challenges associated with obtaining reliable and accurate results through standard MCNP calculations, the investigation focused on the use of weight windows as a variance reduction technique. The findings revealed that the IAEA method tends to provide conservative results only when the same conditions were replicated in the MC simulations. In fact, approximately 50% of the final dose estimated through MC methods accounted for factors that were not considered in the analytical calculations. The primary contributor to scattering (averaging around 30%) was identified as the floor and ceiling. This study underscores the need for caution when relying solely on the analytical approach, as it may not consistently yield conservative outcomes. [ABSTRACT FROM AUTHOR]

The Annular Core Research Reactor (ACRR) Monte Carlo N-Particle (MCNP) model is used by ACRR reactor operators and experiment designers at Sandia National Laboratories for a variety of computational calculations ranging from reactor kinetics parameter estimates and safety analyses to experimental planning. To understand the dominant source of uncertainty within the MCNP model, perturbations in temperature were applied to individual ACRR MCNP fuel rods. Fuel rod temperatures were randomly sampled from a uniform distribution from operational temperatures to quantify temperature-related uncertainty effects. Stochastic mixing was used to blend the cross sections of the desired temperatures using the MCNP continuous and Thermal Neutron Scattering Treatment [S(α,β)] libraries in ENDF/B-VII.1. This uncertainty analysis produced a 640 row × 640 column correlation and covariance matrix of the neutron energy spectra. Positive covariance was produced around the 1-MeV region and the 0.2-eV region. Correlation was found in the thermal and fast energy regions, but no correlation was observed in the slowing-down energy region because interactions in this region are not dominated by fuel. [ABSTRACT FROM AUTHOR]

A coaxial type HPGe detector, and two lanthanide scintillation type detectors have been modeled and characterized by means of Monte Carlo method implemented using the MCNP code. The isotope of interest is 88Rb which decays with energies up to 5 MeV, and can in theory aid in extending the efficiency calibration of the installations of interest in higher gamma rays energies. An 152Eu point source was used for the experimental validation of the MCNP models and excellent agreement was observed between the experimental data and the simulated data of both 152Eu and 88Rb point sources. After the experimental validation of the model, the efficiency calibration was extended for three different source-to-detector distances and the results are presented in this work. [ABSTRACT FROM AUTHOR]

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The nuclear industry produces large quantities of low, intermediate, and high levels of radioactive waste, all of which require safe management during both transport and storage. This study evaluates the radiation shielding effectiveness and mechanical properties of four distinct container materials: Pb Composite Glass, 0.5 Cement-0.5 Bitumen, Concrete (Steel-Magnetite), and C9 (BCBV0.5) Vanadium Oxide-Glass. Using Monte Carlo simulations and theoretical methods, we determined the Transmission Factors (TF) and Half-Value Layers (HVL) for each material. The TF indicates the effectiveness of a material in attenuating radiation, calculated by the ratio of gamma rays exiting the material to those entering it. Lower TF values signify better radiation shielding. The HVL is the thickness of material required to reduce the intensity of gamma rays by half, with lower HVL values indicating more effective shielding. Concrete (Steel-Magnetite) demonstrated superior performance with the lowest TF values (e.g., 1.0 × 10-1 at 0.662 MeV and 1 cm thickness) and HVL values (e.g., 2.5 cm at 1.3325 MeV), alongside a high elastic modulus of 163.15 GPa, indicating its robustness for high-energy gamma-ray applications. Pb Composite Glass also showed strong performance with a TF of 9.5 × 10-2 at 0.662 MeV and 1 cm thickness, an HVL of 2.0 cm at 0.662 MeV, and an elastic modulus of 41.54 GPa. The C9 (BCBV0.5) Vanadium Oxide-Glass, with an elastic modulus of 73.79 GPa, outperformed the 0.5 Cement-0.5 Bitumen mixture in both TF (e.g., 1.15 × 10-1 at 0.662 MeV and 1 cm thickness) and HVL (e.g., 4.2 cm at 1.1732 MeV) measurements, highlighting its potential as a more effective alternative. It can be concluded that C9 (BCBV0.5) Vanadium Oxide-Glass presents promising properties for future advancements in radiation protection, warranting further research and optimization.

Limited by the characteristics of the boron neutron capture therapy (BNCT) treatment beam, it is difficult to measure the γ ray energy spectrum of the source term. There are few relevant work has been carried out. Up to now, only the γ energy spectrometer directly measured the source term γ rays, but the spectrometer was not been calibrated and unfold. In order to obtain an accurate γ energy spectrum of the source term, it is necessary to calibrate the γ energy spectrometer and establish the reliable response function of the system. This paper briefly introduces the γ energy spectrometer designed for BNCT treatment beam, and mainly introduces the calibration method suitable for this spectrometer. The spectrometer consists of a high purity germanium (HPGe) detector and a shield system. Since the entire spectrometer is designed for high intensity neutron/γ mixed fields, the shield effect of the spectrometer is so effective, and it is impossible to calibrate the γ response of the entire spectrometer system through experimental methods. By analyzing the spectrometer system, the dead layer distribution of the HPGe detector is the main influence factor, and other structures can be accurately obtained through X-ray radiography and geometric measurements. Therefore, the calibration method of the spectrometer is as follows: obtain an accurate model of the HPGe detector by X-ray radiography and calibration; use precision measurements, fine geometric measurement and material analysis to obtain an accurate model of the shield, and establish a calculation model of the spectrometer; acquire the response function of the spectrometer system through Monte Carlo N particle transport code (MCNP) calculation, and the energy spans from 0.1 MeV to 1 1 MeV with an energy interval of 10 keV. Among them, the HPGe detector dead layer distribution can be confirmed using the following steps. The efficiency of the HPGe detector was calibrated utilizing standard γ sources and high-energy γ reference radiation fields. The dead layer distribution of the HPGe detector was adjusted using MCNP to ensure agreement between calculated and measured efficiencies, thereby determining the geometric distribution of the dead layer. Therefore, using a combination of Monte Carlo simulation and experimental methods, the results can be traced to the corresponding national or national defense standards, ensuring the accuracy and reliability of the key parameters of the spectrometer, and laying the foundation for subsequent spectrum unfold work.

As a new type of energy system, thermoelectric isotope battery system is widely used in deep space and deep sea exploration missions. Strontium titanate fuel is used as the heat source for thermoelectric isotope batteries. The radioactive isotopes of 90Sr and 90Y in the strontium titanate heat source emit β ray, which interact with substances to produce bremsstrahlung radiation. Due to the strong penetrating power of bremsstrahlung radiation, it is necessary to install a shielding layer on the surface of the isotope radioactive source strontium titanate to reduce the radiation dose of the thermoelectric isotope batteries. When the thickness of the shielding layer exceeds a certain value, the statistical error of the Monte Carlo method increases significantly. Variance reduction technique can be used to improve the accuracy of shielding calculation for thermoelectric isotope batteries. At the same time, the direct electron transport simulation takes a long time, which is much longer than the photon simulation, and the photons produced by the electrons can be processed by the thick target bremsstrahlung (TTB) model, thus improving the computational efficiency. In this paper, the variance reduction technology of cosRMC program and the application of TTB radiation model in the shielding calculation of isotope batteries were studied and verified. Based on the design scheme of SNAP-21 structure in the United States, the thermoelectric isotope battery is modeled by cosRMC program. The influence of depleted uranium on the surface dose rate was analyzed by changing the thickness of the radial and axial outer shielding layers of the thermoelectric isotope battery, and the calculated results were compared with those of MCNP program. The comparison of the two programs shows that the statistical error of dose rate of depleted uranium shield surface calculated by cosRMC and MCNP is in good agreement with ±3σ range when different thickness shielding layers were used in radial and axial directions of thermoelectric isotope batteries. In order to reduce the statistical error of the calculated results, the cell importance variance reduction was used. The results show that the statistical error of the cosRMC and MCNP programs decreases at the radial 5 cm and axial 4 cm of the battery by 52.515% and 39.338%, respectively. The dose rate relative errors calculated by cosRMC and MCNP are in good agreement within the range of ±3σ, which further verifies the accuracy of cosRMC program in the calculation of thermoelectric isotope battery shielding.

This study investigated the radiation shielding performance of the xBaO-(30-x)TeO2-35Bi2O3-35B2O3 glass system, where x = 5, 10, 15, 20 and 25, using Phy-X/PSD software and MCNPX Monte Carlo code. Key radiation shielding indicators, including attenuation coefficients, half value layer (HVL), tenth value layer (TVL), effective atomic number (Zeff), mean free path (MFP), and build-up factors, were measured within the energy range of 0.015–15 MeV. Further, the study assessed the glasses' capacity to block fast neutrons, using the fast neutron removal cross-section (ΣR) as a metric. The results showed that gamma shielding effectiveness follows the order: BaTBiB-1 < BaTBiB-2 < BaTBiB-3 < BaTBiB-4 < BaTBiB-5 at all photon energies. Additionally, the ΣR values for BaTBiB-1 to BaTBiB-5 glasses were 0.10179, 0.10263, 0.10362, 0.10581, and 0.10623 cm−1, respectively. Therefore, the findings demonstrate that including BaO enhances the gamma and neutron shielding characteristics of tellurite glasses, making them valuable for radiation-related applications.

VERA benchmark cases have been introduced by ORNL and detailed guidelines have been provided including the burnup chain data for a depletion analysis. The purpose of this paper was to analyze the performance of MCNP Monte Carlo code using VERA benchmark cases. Research on the VERA benchmark cases become important because its results will be useful for designing SMRs planned in the country. The VERA 2B and VERA 2P were selected as a representation for a typical fuel assembly of LWR configurations. The kinf MCNP calculations show a good agreement with the MCS and STREAM predictions at the Beginning of Cycle (BOC), Middle (MOC), and End (EOC). The kinf value at BOC calculated using MCNP for VERA 2B case shows a good agreement compared to KENO, Serpent, and OpenMC with a difference of less than 60 pcm while for VERA 2P case with a difference of less than 90 pcm. Doppler Temperature of reactivity Coefficient (DTC) and Moderator Temperature (MTC) results were negative and both became more negative as fuel burnup increased. The βe f f values of both cases were close to 670 pcm during BOC since both cases use 235U fissile nuclear material. The 239Pu production in VERA 2B was higher than 2P since higher 238U in 2B and 24 pins of mixed lower enrichment and gadolinium in 2P lowers its heavy metal loading followed by the neutron spectrum shifting. It can be concluded that these results can be used in the development of computational performance for overall core analysis to strengthen the basic design of the SMR to be built in Indonesia. [ABSTRACT FROM AUTHOR]

AbstractExternal exposures to airborne radioactivity at nuclear power sites are typically based on semi-infinite clouds with uniform concentrations and corresponding effective dose rate coefficients (EDRCs), along with the simplifying assumption that the radioactive clouds extend to infinity around the receptor of interest. The two regulatory models that are typically employed for finite-cloud adjustments to submersion doses, namely, Regulatory Guide 1.183 for hemispherical clouds and the International Commission on Radiological Protection (ICRP) document ICRP-30 for spherical clouds, were purposely oversimplified to facilitate their implementation. As a result, dose projections can be significantly underestimated under certain circumstances, particularly with radionuclides emitting low-energy photons and/or particles. In addition, these adjustments do not account for scatter radiation off surrounding walls, ceilings, and floors of typical occupational settings.In recognition of these limitations and for the mitigation thereof, Veinot et al. published an article [Rad. and Environ. Biophy., Vol. 56, p. 453 (2017)] that provided monoenergetic photon, electron, and positron EDRCs based on elaborate Monte Carlo computations for submersion in three typical occupational settings at nuclear facilities (namely, an office, a laboratory, and a warehouse). Included in the article were also EDRCs for exposure to 45 noble gases airborne within the said occupational settings. However, extrapolation of the results to other settings was limited due to geometry variations.Even so, as described in the present paper, the Veinot et al. article provided the basis for the definition of a straightforward model for extrapolation of the Monte Carlo–derived EDRCs to other occupational settings and radionuclides. The objective of the EDRC extrapolation model in this paper is to provide a simple but comprehensive approach to licensing-basis submersion dose calculations consistent with the recommendations in ICRP Publication 103.The model is reasonably accurate and applicable to submersion volumes that are smaller than an office and larger than a typical warehouse, as would be needed for control room habitability evaluations. It is emphasized that the model is only suitable for the external dose computation of airborne noble gases and particulates; dose contributions via the inhalation pathway and from direct shine from contaminated surfaces need to be evaluated separately.The EDRC extrapolation model was programmed into an Excel workbook (OccuSetEDRCs-R0.xlsx) that is available to interested parties; see Supplementary Data section for further details. The model can handle all 1252 radionuclides in ICRP-107 and submersion volumes ranging between about 40 and 4000 m3, with an upper estimated error of about 10% on average for large submersion volumes. [ABSTRACT FROM AUTHOR]

AbstractThe Molten Uranium Breeder Reactor (MUBR) is a radically new reactor concept with a mixed-energy spectrum. MUBR is fueled with molten uranium metal in large-diameter fuel tubes and is cooled by circulating molten uranium fuel through a heat exchanger. The reactor has heavy water as moderator, and the reactivity of the reactor is primarily controlled by the voiding effect of the moderator through an innovative control cavity structure design. Because the MUBR design is vastly different from most existing fission reactors, neutronics analysis must be performed for many different combinations of design parameters to identify viable and optimum design configurations. To facilitate the neutronics analysis, a proprietary program called MUBR6gen is being developed to provide a pipeline tool to expedite the process. MUBR6gen employs two well-established neutronics codes, i.e., MCNP and SCALE, to perform standard neutronics calculations for MUBR by automating input preparation and output processing. In addition, MUBR6gen ensures consistency of the MCNP and SCALE inputs and compares the outputs of the two codes to warrant the simulation results. Augmented with MUBR6gen, standard neutronics analysis was carried out on a small-scale MUBR design, which serves as a model problem in the paper. The neutronics performance characteristics of the model reactor were obtained and discussed in a code-to-code pattern. An overall very good agreement between the results of the two neutronics codes was established. Based on the success of the model problem analysis, further neutronics analysis using MUBR6gen was extended for a set of MUBR variant designs. Meaningful and promising fuel cycle analysis results for the 10 different designs were achieved and discussed. These results are used to identify the best MUBR candidates in terms of fuel lifetime and utilization efficiency for future applications. [ABSTRACT FROM AUTHOR]

AbstractAn accurate assessment of the deposited energy across a reactor geometry allows for a better determination of the heat removal requirements and ensures effective cooling after shutdown. This work discusses several methods in detail that target the heat deposition in specific reactor regions, leading to a choice of a prevalent method for the calculation workflow. This paper introduces a delayed heating calculation workflow based on the formal three-step process. The workflow relies on the MCNP-ORIGEN Activation Automation tool for performing the first two steps of the process, while the third step is conducted via MCNP photon transport simulations. This paper showcases two applications to demonstrate the workflow and simulation outputs. These include an Advanced Test Reactor (ATR) experiment and the RA-6 reactor structures. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The estimation of flux in radiation transport Monte Carlo problems needs to calculate the volumes and surface areas of the geometric regions. The particle flux is often estimated as the track length per unit volume or the number of particles crossing a surface per unit area in Monte Carlo transport problems. Various representations such as constructive solid geometry (CSG), boundary representation (B-Rep), and combinatorial geometry (CG) are proposed in the literature for geometry modeling and calculation of surface area and volume. MCNP series and OpenMC as Monte Carlo particle transport codes utilize CG modeling and are not able to calculate surface area as well as volume for non-rotationally symmetric or non-polyhedral cells. In this work, a comprehensive approach based on the Cauchy-Crofton formula using the Monte Carlo method has been implemented to the radiation transport codes as an extra module for computing surface area and volume of complex geometries. We used a random sampling procedure to create the required probe lines and points in the computational approach. The results show that this method can accurately compute surface areas and volumes of complex geometries with a relative error of less than 0.1% and a short computation time of a few seconds, which is not achievable with the cuurent MCNP and OpenMC modules. [ABSTRACT FROM AUTHOR]

The Joint European Torus (JET) is playing an important role in preparing for the operation of the future world's largest tokomak, ITER. In this respect, the tritium campaign (C40) and second deuterium–tritium experiment (DTE2, C41) took place in the JET during the years 2021 and 2022. In this work, a corresponding irradiation scenario was utilized for the activation calculations of eight material foils located at the JET outer long-term irradiation station (OLTIS). Neutron-induced activities and dose rates at a 30 cm distance after shutdown at specified cooling intervals were calculated with the FISPACT-II code, employing the EAF-2010 nuclear and TENDL-2021 data libraries. The Monte Carlo MCNP6.2 particle transport code equipped with the FENDL-3.1d nuclear data library was used for the calculation of the neutron flux densities. [ABSTRACT FROM AUTHOR]

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

In multiplying systems, such as nuclear reactors and criticality experiments, it is desirable to place the fissile material in the optimal or 'best' way to reduce the critical mass of the system as well as to achieve uniform fuel burnup. This paper considers two methods, namely Pontryagin's maximum principle (PMP) and Monte Carlo (MC) perturbation for estimating a minimum critical mass configuration. These methods are applied to an elementary multizone model of a pressurized water reactor (PWR) and a criticality experiment to estimate the minimum critical mass. It is found that while twogroup diffusion theory with PMP predicts a minimum critical mass, more detailed MC simulations with MCNP5 show a consistent reduction in critical mass when fissile fuel is placed in inner zones. Such a distribution reduces the fissile material requirement but is undesirable due to the higher power peaking. MC simulations show that for a three-zone model of the KORI 1 PWR, a uniform fissile distribution gives criticality for 1.09 atomic percent (at.%) enrichment, whereas non-uniform fissile distribution (0.6, 1.6, 0.6 at.%) reduces the critical mass by 14%. The changes found from MC simulations were subsequently predicted from first-and second-order derivative sampling. It was found that substantial computational savings can be achieved for large-scale optimization problems. In the case of a criticality experiment, MC derivative sampling was also used to estimate optimal fissile distribution for minimizing the critical mass. [ABSTRACT FROM AUTHOR]

Thermoluminescence dosimeter (TLD) is widely used as a personal and medical dosimeter. Several TLD materials show the characteristics of mass energy absorption coefficient and energy response relative to ICRU (International Commission on Radiation Units and Measurements) issue material as an equivalent material for human body soft tissue. This research aims to analyze the effect of Dy3+ dopant on the mass-energy absorption coefficient and relative energy response of Lithium Magnesium Borate (LMB) materials. The simulation was carried out using Monte Carlo N-Particle (MCNP) software. Calculations based on simulation and theoretical results will be compared statistically using paired t-tests. The study showed that adding a Dy3+ dopant to TLD material made of Lithium Magnesium Borate (LMB) only affected the mass-energy absorption coefficient and relative energy response for low radiation energy. Adding Dy3+ dopant increased the mass energy absorption coefficient and relative energy response in a reasonably small value. Based on these results, LMBDy3+ produces a better mass-energy absorption coefficient value for TLD materials. The results of the statistical tests show a significant difference in the mass energy absorption coefficient value. At the same time, there is no significant difference between the simulation results and theoretical calculations for the relative energy response.

To obtain cobalt target activity data after a cycle of irradiation in a pressurized water reactor, the cobalt target dose distribution was measured underwater using alanine dosimeters. The effects of cobalt target structure, water medium, measurement device, operation, and environment were corrected using Monte Carlo calculations. The feasibility of the method and device was confirmed through both experimental and Monte Carlo comparison. The underwater measurement data of cobalt target activity matches the theoretical calculated values, with an uncertainty of 6.5%.

90Sr has strong bremsstrahlung radiation, before manufacturing the radioisotope heat source, it is necessary to first simulate the composition of the 90SrTiO3 ceramic pellet, design the structure of the fuel cladding, and evaluate the bremsstrahlung radiation of the pellet and fuel cladding according to the manufacturing process. In order to reduce the exposure dose of thermoelectric conversion devices, the article based on β shielding principle has proposed three methods to reduce bremsstrahlung radiation from radioactive sources. The MCNP simulation results show that using aluminum foil to wrap the core blocks, optimizing the core block ingredients, and adding boron nitride gaskets can effectively reduce bremsstrahlung. All three methods will change the design size of the fuel cladding, and the engineering feasibility of wrapping aluminum foil is poor. The research results will guide the design and manufacturing of the source part of the Sr-90 thermoelectric generator, and provide theoretical methods for radiation optimization of the generator.

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark.

To organize the safe handling of activated material, knowing the residual dose rates is crucial. In this work, we present the pre-experiment activation analysis for an experiment in which tungsten blocks are irradiated by 800-MeV protons. In this analysis, we use the Monte Carlo N-Particle (MCNP) code for radiation transport, Attila4MC for unstructured mesh generation, and Activation in Accelerator Radiation Environments (AARE), including CINDER2008, for activation analysis. If the tungsten blocks must be removed within a day after the experiment, then exposure to personnel entering the room must be reduced. One exposure-reduction strategy is to add carbon steel shielding around the tungsten blocks, efficiently reducing the dose from the activated tungsten. However, the shielding becomes activated itself during irradiation: 56Mn is the dominant contributor for short decay times. The actual schedule at the time of the experiment allowed sufficient cool-off time for the tungsten in the room so that additional shielding was not necessary. A less rigorous comparison of the calculated values with the post-experiment measurements showed reasonable agreement. [ABSTRACT FROM AUTHOR]

The effects of both proton and neutron nuclear data libraries on the neutron field produced in light targets and transported through shielding material were assessed. The general-purpose nuclear data libraries JENDL-4.0/HE, TENDL-2017, ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.1.2, and JEFF-3.3 were benchmarked using the MCNP6.2 Monte Carlo transport code. Three experiments from the SINBAD database, namely, 52-MeV protons hitting a carbon target and 43- and 68-MeV protons hitting a 7Li target, were selected for benchmarking. This selection was made according to their relevance for the MYRRHA accelerator-driven system to support the design of beam dumps and shielding of the 100-MeV section of the MYRRHA accelerator. It is demonstrated that the prevailing factor determining the transmitted neutron spectrum shape is the proton library from which the secondary neutrons are sampled. The choice of neutron library applied for the secondary neutron transport in the carbon and lithium targets, concrete and iron shielding, is of second-order significance. [ABSTRACT FROM AUTHOR]

The European Spallation Source (ESS) is currently under construction in Lund, Sweden. At the time of the first beam-on-target, foreseen in 2025, a first set of neutron scattering instruments will be commissioned, in addition to a test beamline for characterization of the source. For every planned beamline, all optics components will already have been installed in the monolith shielding surrounding the spallation target, but instruments will still be under construction in the neutron bunker area. Blocks of shielding, referred to as temporary beam stops or wall plugs, are then required to capture neutrons that will come out through the bunker wall feedthroughs during operation. This paper presents the results of parametric studies carried out to design temporary beam stops to assess the possibilities of reducing the weight and cost of such components and also to investigate the shielding efficiency of a wall plug as an alternative solution. [ABSTRACT FROM AUTHOR]

A virtual reality (VR) system for Monte Carlo (MC) transport simulation codes, MCNP and Particle and Heavy Ion Transport Code System (PHITS), was developed. This system is aimed to provide a more reliable confirmation of the simulation geometry through an intuitive user interface. The VR system is implemented by extending the Geometry and CROSS Section VIEWer (Gxsview) code to run in a JavaScript environment, and it takes advantage of web technology and does not require dedicated hardware or geometry information supplied by computer-aided design software. Therefore, only WebXR-compatible head-mounted displays and browsers are required on the user side. In the VR space, visualization of the calculation geometry, cell selection, and cell grabbing operations is possible with a handheld controller. Further enhancements will be made in the future, such as performing shielding calculations with the cell changed by grabbing operations. [ABSTRACT FROM AUTHOR]

Copyright of Journal of Isotopes is the property of Journal of Isotopes Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Journal of Isotopes is the property of Journal of Isotopes Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Research and development of the DEMOnstration power plant (DEMO) breeder blanket (BB) has been performed in recent years based on a predefined DEMO tritium breeding ratio (TBR) requirement, which determines a loss of wall surface due to non-breeding in-vessel components (IVCs) which consume plasma-facing wall surface and do not contribute to the breeding of tritium. The integration of different IVCs, such as plasma limiters, neutral beam injectors, electron cyclotron launchers and diagnostic systems, requires cut-outs in the BB, resulting in a loss of the breeder blanket volume, TBR and power generation, respectively. The neutronic analyses presented here have the goal of providing an assessment of the TBR losses associated with each IVC. Previously performed studies on this topic were carried out with simplified, homogenized BB geometry models. To address the effect of the detailed heterogeneous structure of the BBs on the TBR losses due to the inclusion of the IVCs in the tokamak, a series of blanket geometry models were developed for integration in the latest DEMO base model. The assessment was performed for both types of BBs currently developed within the EUROfusion project, the helium-cooled pebble bed (HCPB) and water-cooled lead–lithium (WCLL) concepts, and for the water-cooled lead and ceramic breeder (WLCB) hybrid BB concept. The neutronic simulations were performed using the MCNP6.2 Monte Carlo code with the Joint Evaluated Fission and Fusion File (JEFF) 3.3 data library. For each BB concept, a 22.5° toroidal sector of the DEMO tokamak was developed to assess the TBR and nuclear power generation in the breeder blankets. For the geometry models with the breeder blanket space filled only with blankets without considering IVCs, the results of the TBR calculations were 1.173, 1.150 and 1.140 for the HCPB, WCLL and WLCB BB concepts, respectively. The TBR impact of all IVCs and the losses of the power generation were estimated as a superposition of the individual effects. [ABSTRACT FROM AUTHOR]