Shielding Calculations Research Articles

Comprehensive assessment of broad beam transmission factors for nuclear medicine shielding.

This study aims to calculate broad beam transmission factors (BBTF) for common nuclear medicine shielding materials (lead, lead glass, concrete, barite concrete, and iron) using consistent Monte Carlo (MC) simulations. Additionally, the feasibility of fitting these transmission factors to the Archer model was explored. Photon emission spectra and biodistributions of eleven radionuclides (11C, 15O, 18F, 67Ga, 68Ga, 99mTc, 123I, 131I, 153Sm, 177Lu, and 201Tl) were analyzed. BBTFs were calculated for five shielding materials using both point sources and anthropomorphic phantoms to reflect practical clinical conditions. Validation was performed by comparing results with published data. BBTFs for eleven radionuclides and five shielding materials were obtained. The calculated half-value layer, tenth-value layer, centesimal value layer, and millesimal value layer thicknesses demonstrate good agreement with experimental and reference data for point sources. Results for phantom sources extend existing methodologies for PET shielding calculations to other radionuclides used in clinical practice. Findings indicate a reduction in shielding thickness due to changes in the radiation spectrum from body self-attenuation and highlight the impact of radionuclide biodistribution on BBTFs. The Archer model was not suitable for all sources, particularly low-energy ones. The study provides a validated basis for BBTF calculations across different sources and materials. Results demonstrate the need to account for body self-attenuation and biodistribution in shielding design. Future work should consider different biodistributions, phantom positions and generalized fitting models.

Systematical Theoretical Study of the Nonpoint Source Effects in Nuclear Medicine Shielding Calculation

Abstract The objective of this note was to systematically study the effects from nonpoint radiation sources on the nuclear medicine imaging room or uptake room shielding design. The dose from a nonpoint source to a location at a certain distance to the center of the source was calculated. A correction factor could be applied to the dose calculated using a point source model with the same distance to compensate the nonpoint source effects. The correction factor was a function only depending on the source relative dimension, the ratio of the source’s geometric dimension to the distance. Theoretical and numerical calculation were performed to calculate the correction factor on a one-dimensional line source, two-dimensional circular disc source, three-dimensional sphere, and cylinder sources. Our study indicated that for large nuclear medicine imaging rooms, the point source model was a good approximation for shielding design. For a small uptake room, a nonpoint source correction may be considered to calculate the barrier thickness.

Analysis of the NMR Parameters Shielding and Spin-Spin Coupling Constants of Glycine Conformers.

In this study, we worked at the CCSD/aug-cc-pVTZ level to obtain the conformers of glycine in its neutral and zwitterionic forms in the gas and water phases. We then computed the NMR properties (spin-spin coupling constants and nuclear magnetic shieldings) at the SOPPA/aug-cc-pVTZ-J level. We attempt to elucidate the apparent discrepancy arising from two previous works by Valverde et al. [J. Chem. Phys., 2018, 148, 024305] and Caputo and Provasi [Sci, 2021, 3, 41]. Our optimized structures align with previous theoretical predictions, although some geometries were not found. Additionally, our results suggest that in the gas phase the Δ 1J(O,C) = 1J(Ocis,CO) - 1J(Otrans,CO) is positive when the acid group is in the trans conformation and negative in the cis conformation; however, this trend is not well reproduced in water. From magnetic shielding calculations, we cannot distinguish between an O-H in the cis or trans conformation.

Optimization progress of large-scale radiation shielding Monte Carlo simulation software based on AIS variance reduction technique system: MCShield

The development of novel nuclear facilities has brought safer and more efficient energy options but also poses significant challenges for radiation shielding calculations. MCShield, developed by the Radiation Protection and Environmental Protection Laboratory at Tsinghua University, is a Monte Carlo program designed for coupled neutron/photon/electron transport in radiation shielding calculations. It incorporates a system of variance reduction techniques based on Auto-Importance Sampling (AIS) to address the deep penetration problem commonly encountered in the field of radiation shielding. The accuracy and computational efficiency of MCShield have been validated through benchmark problems and real-world applications. However, the current AIS system faces limitations in complex scenarios, user-friendliness, and reliance on user experience. To address these issues, we optimized the size, shape, and energy parameters for the AIS variance reduction method, expanded the use of regular virtual surfaces, and introduced an irregular AIS virtual surface method. Additionally, we developed an automatic generation method for AIS virtual surfaces and implemented automatic calculation for these surfaces. AIS Energy Bias Method was proposed to improve convergence across different energy intervals. These improvements enhance the applicability and refinement of the AIS virtual surface parameters, significantly boosting the overall performance of MCShield.

Effect of the oblique incidence of radiation beams on emerging radiation behind lead and concrete shields: a multilayer method for dose transmission calculations

Objective. For calculating shielding in x-ray rooms, it is often assumed that the beams impinge perpendicularly on the protective barriers. This is not always true, but this premise simplifies the calculations and enhances protection by being a conservative calculation. In this work, a method for calculating radiation transmission through planar shielding that considers the obliquity of the incident beam is presented.Approach. The output of the method produces energy spectra according to the direction of radiation impinging on the shielding. Four angles of incidence on the barrier are considered, along with monoenergetic pencil beams with energies ranging from 10 to 150 keV and two materials: lead and concrete. The direction of emerging photons is discretized into 49 different direction vectors. Monte Carlo calculations are performed for thicknesses of 0.1, 0.5, and 1.0 mm of lead, and 1, 5, 10, and 15 cm of concrete. Additionally, a multilayer iterative method is implemented for calculating attenuation of other thicknesses.Main results. The distribution of radiant energy according to the coordinates of its directional vector illustrates the effect of the obliquity of the incidence and the significance of the shielding material employed. In the case of concrete, the dispersion of radiation away from the original direction of incidence is much more pronounced than in the case of lead at energies below its K-edge. The multilayer iterative method provides highly accurate values of transmitted radiant energy in both monoenergetic and polyenergetic beams, for both lead and concrete, across the various studied incidence directions.Significance. Considering the direction of the photons reaching a shield and the direction of the photons passing through it allows multilayer composite shielding calculations to closely approximate the calculation made for the composite shielding.

The curious case of S4N4 − A reinvestigation

This work employs quantum chemical methods to explore electronic properties and bonding of the S4N4 molecule. We utilized calculations of multidimensional off-nucleus magnetic shielding, anisotropy of the induced current density (ACID), natural bond orbitals (NBOs), and atoms in molecules (AIM). The unique S4N4 system comprises bent bonding, absence of NN, weakness in the centers of SS bonds, and polar SN bonds. Our findings uncover coexisting of both electron localization and delocalization in S4N4. The magnetic analysis discloses curved connections between the nitrogen atoms and the SS bonds, drawing possible electron delocalization pathways as two pyramid-like regions of magnetic activity. Our magnetic calculations were performed at DFT/GIAO-B3LYP using Jorge-TZP-DKH and 6-311++G(2d,2p) basis sets. No remarkable qualitative differences were noticed between the obtained magnetic features from these two basis sets. Capturing the magnetic details associated with structural features shows superiority of ZZ component of the magnetic shielding tensor (σzz(r)) over the isotropic magnetic shielding (σiso(r)).

Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors.

Analysis of transport-activation internal coupling method implemented in Monte Carlo code cosRMC

When fusion nuclear reactor is in operation, neutron activation causes a large number of radioisotopes. Activation calculation is a very important step in both of reactor shielding calculation and radiation safety analysis. The capability of transport-activation coupling calculation was developed in Monte Carlo code cosRMC though the built-in burnup solver Depth. In the process of neutron transport, the one-group cross section of activation-related was calculated by internal coupling method and the manipulation and transfer of input file for energy spectrum to external activation code was no longer necessary, which is more efficient and flexible than traditional external coupling. A cuboid iron model is established, and the activation calculation is performed by cosRMC and ALARA respectively. In the constant energy spectrum mode, the relative error of 56Fe atomic density between two codes is 0.0019 %, which verified the accuracy of capability of activation internal coupling calculation in the cosRMC code. The relative errors of 57Fe and 58Fe atomic density are 17.97 % and 36.75 % between constant and variable energy spectrum. This showed that whether energy spectrum remains constant has a significant impact on the results. Moreover, the one-group reaction rate with continuous energy/ multi-group cross section were different, which is because the pre-generated multi-group cross section libraries are independent of the geometry and energy spectrum of the specific problem. The resonance shielding effect of the specific problem cannot be accurately considered when using the multi-group cross section, resulted in relative errors of 60 % for the one-group cross section of 56Fe(n,γ) reaction. cosRMC can directly use the continuous energy method to calculate the problem dependent one-group section, which will give more accurate results.

Neutron shielding calculation for DEMO-Prad/SXR measurement system

In plasma fusion devices, the selection of shielding materials is one of the challenges for plasma diagnostic and control systems under high radiation levels during long operation times. This study presents the effect of shielding materials on neutron flux for a radiated power and soft x-ray core intensity measurement system for a DEMOnstration power plant. A calculation model of neutron shielding was used to investigate the neutron shielding performance of borides and carbides for a large distance from plasma to the diagnostic system. The neutron fluxes were characterized for three points close to the measurement system location. The related neutronic calculations were performed with an ADVANTG hybrid code to obtain neutron flux distribution and attenuation rate depending on the thickness of shielding materials. The results indicate that B4C, W2B5, and WB4 are the most effective options to serve as shielding material due to the effect of boron on neutron shielding effectiveness.

Monte Carlo simulation of air kerma rate constants for different radionuclides

The air kerma rate constant values for different radionuclides are often used in radiation shielding calculations, calibration of instruments, estimating the absorbed dose from brachytherapy sources, etc. Usually, air kerma rate constant values are calculated in a deterministic way. This work evaluates the possibility of air kerma rate constant estimation through Monte Carlo simulations. Ten different radionuclides were considered in this preliminary study, covering low and high-energy gamma spectra. Mean percentage differences of less than 10% relative to the results of reference works were found. Most of our results present less than a 5% relative percentage difference considering the literature results. The air kerma rate constants calculated in this work agree with reference results, thus validating the methodology, the photon spectra database used, and the Monte Carlo simulations with EGSnrc. Future work can extend the calculation of air kerma rate constants for other radionuclides and study the influence of different sources of uncertainty in its results.

Simple approach to evaluate safety requirements to establish nuclear cardiology unit: Shielding and Occupational dose calculations

Simple approach to evaluate safety requirements to establish nuclear cardiology unit: Shielding and Occupational dose calculations

Radiation shielding analysis and criticality safety calculation for spent fuel transportation cask

China anticipates a shortage in storage capacity for spent fuels at reactor sites for the AP1000 NPPs within the next 10 to 20 years. We proposed a conceptual design for a transportation cask capable of loading 21 spent fuel assemblies in AP1000 reactors and studied its criticality safety and radiation safety characteristics. Throughout these operations, three key concepts were taken into account: source term, radiation shielding, and criticality calculations. The sources of the 4.95% enriched spent assemblies following a cooling period of 21 years were calculated by the cosNAB code. The results of the source term computation were utilized as input data for the Monte Carlo code cosRMC, which was employed for shielding calculation and criticality safety analysis. The calculation results have verified the feasibility of the design, which can provide a reference for the development of spent fuel transportation casks for the AP1000 reactor.

A weight window approach for dose rate calculations in Lead-cooled Fast Reactors with OpenMC

This paper presents the application of a variance reduction technique for radiation shielding calculations in Lead-cooled Fast Reactors (LFRs) using the OpenMC Monte Carlo code. The study focuses on assessing the computational efficiency of the Method of Automatic Generation of Importances by Calculation (MAGIC). This technique is employed to generate weight windows for Monte Carlo simulations, aiming at enhancing the efficiency of dose rate estimations for LFRs. Through a detailed study of a LFR mock-up model, the performance differences between a MAGIC-accelerated and a non-accelerated simulation are investigated in terms of both accuracy and computational cost. The results demonstrate the benefits of the variance reduction method, showing its effectiveness in spreading particles throughout the model while improving the accuracy of Monte Carlo estimates far from the core. In addition, this work demonstrates the performances of the MAGIC method in eigenvalue calculations. This work shows an interesting optimisation of Monte Carlo simulations for LFRs and can be considered as a first milestone for newcleo’s radiation shielding studies.

Эмнэлгийн рентген төхөөрөмжийн цацрагийн хамгаалалтын тооцоолол

In this study, we performed radiation shielding calculations for the diagnostic X-ray equipment at Rekha Med Hospital. The lead thickness required for radiation protection, as determined by the methodology outlined in the National Council on Radiation Protection and Measurements (NCRP) Report No. 147, was approximately 2 mm. Air kerma measurements were taken outside the X-ray room to validate the shielding effectiveness with a 2 mm thick lead barrier installed along the walls. The results confirmed that the air kerma dose did not exceed the regulatory dose limit of 0.5 mSv/H in uncontrolled areas, indicating sufficient shielding.

Characterization of neutronic parameters and radiation shielding design for an aqueous homogeneous reactor

In this research, an aqueous homogeneous reactor is simulated and its neutronic parameters are investigated. Due to the negative reactivity caused by temperature change and void formation, the cold excess reactivity is considered equal to 4 βeff. Also, calculations of radiation shielding and neutronic characterization of the reactor are performed. For different heights of uranyl sulfate fuel, the effective multiplication factor (keff) of the reactor is calculated under cold zero power and hot full power conditions to assess the required excess reactivity. The amount of fuel burnup and changes in the keff are investigated for 500 days, finally 200 operating days are chosen as the working duration of the reactor. The control rods worth (CRW) for different diameters of boron carbid and reactor safety parameters including shot down margin (SDM) and safety reactivity factor (SRF), temperature and void coefficients of reactivity are analyzed. In order to design a suitable radiation shield for this reactor, two different shields are considered. The first shield is a combination of polyethylene thicknesses and barite concrete, and the second shield is made only of barite concrete. The total thickness of the first and second shield was 230 and 180 cm, respectively. Due to the problems of using polyethylene (e.g. problems related to machining, shield dimensions and low melting temperature), the second shield is chosen as the suitable radiation shield for this reactor. Using this radiation shield, the total dose rates of neutron and gamma rays satisfy the ICRP dose limits.

Estimation of weight window parameters based on recursive Monte Carlo approach for reactor shielding problems

Weight window, a variance reduction tool, is often used to improve the performance of radiation shielding calculations. One common issue with the weight window is determining the optimal set of weight window parameters for solving deep penetration shielding problems. To address this issue, the recursive Monte-Carlo methodology has been used with the in-house TRAWEI code. The program is responsible for generating the optimal weight parameters in a single run with minimum computational time. This paper presents the results of a numerical test conducted using a simple reactor model to evaluate the performance of the developed weight generator program. The findings reveal that MCNP simulations utilizing TRAWEI-generated weight values exhibit significantly higher calculation efficiency compared to both analog simulation and MCNP simulation using weights generated by the existing MCNP weight window generator. Overall, the utilization of the RMC methodology has shown its potential to significantly contribute to deep penetration shielding calculations.

Addendum to Effective Dose Transmission Ratio of Photons of 85 Radionuclides for Shielding Calculation of Radiation Facilities—Additional Data to “Data for Shield Calculation of Radiation Facilities (2015)”—

Addendum to Effective Dose Transmission Ratio of Photons of 85 Radionuclides for Shielding Calculation of Radiation Facilities—Additional Data to “Data for Shield Calculation of Radiation Facilities (2015)”—

Critical analysis of the existing approach to the calculation of radiation shielding in X-ray rooms

The paper demonstrates the limitations of the existing methodology for calculating the radiation shielding of X-ray rooms presented in SanPin 2.6.1.1192-03. It is shown that the algorithm for calculating the attenuation coefficient by the barrier does not take into account the features of the attenuation of different components of X-ray radiation: direct, scattered, and leakage radiation, which differ in the region of occurrence, intensity, energy spectrum, and other parameters. Instead, a single formula is used for all components, and the differences in their attenuation are taken into account by a parameter the values of which for different radiation components are in no way justified and are questionable. The calculation also does not take into account the distribution of the workload of X-ray machines by the tube voltage, the significant attenuation of direct X-ray radiation by additional structures necessary for image acquisition. The values of the radiation output of X-ray machines recommended for shielding design in SanPiN 2.6.1.1192-03 are 2–3 times overestimated in relation to the measured values. This leads to an unreasonable overestimation of the requirements for the thickness of radiation shielding in X-ray rooms and accordingly to suboptimal spending on healthcare. It is necessary to develop a new document to replace SanPiN 2.6.1.1192-03.

MAGIC-GPS global variance reduction method for large-scale shielding calculation

We propose a Method of Automatic Generation Importance by Calculation with Growth Population Scheme (MAGIC-GPS) for efficient and accurate shielding calculation. The new method adopts a Growth Population Scheme (GPS) in the reciprocal iteration process of flux and importance for the Monte Carlo fixed-source calculation. Mathematical derivation is performed to prove that the GPS scheme is conducive to a rapid flux distribution convergence and can improve the Figure of Merit (FOM) of the Monte Carlo deep-penetration simulation. The effectivity of the MAGIC-GPS method is tested by a single concrete rod model, the H.B. Robinson-2 reactor benchmark (HBR2) and the China Fusion Engineering Test Reactor (CFETR). The results show that the MAGIC-GPS method can accelerate the Monte Carlo deep-penetration simulation, and the faster the population of particles increases, the better the acceleration. The MAGIC-GPS method is helpful for large-scale shielding calculation.

Improved compact representation method for Monte Carlo phase spaces

Phase space files used in particle Monte Carlo simulations can be large, and thereby awkward to distribute and slow to load into memory. To circumvent this, compact source models and phase space representations have been developed. In this study, the aim is to increase the versatility of a previously presented method, based on principal component analysis whitening, by evaluating several modifications to the method on simulated cone-beam computed tomography projection images. Scaling the phase space components before whitening was seen to give the largest improvement and allowed for accurate modelling of phase spaces with a shifted detector and/or a bowtie filter present, configurations where the original method struggled. The modified method reproduced results using the original phase space to within a few percent both in energy fluence on the detector and in simulated projection and scatter images of a patient pelvis, without obvious position-dependence of the errors. Hence, the modified method was deemed sufficiently accurate for most applications relying on cone-beam computed tomography simulations. The final aim is to use the more versatile method for a broader range of phase space-based applications such as linear accelerator beam simulations and room shielding calculations.