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Radioactive waste resin is generated during the operation of nuclear facilities such as nuclear fuel recycling，reactor cooling water treatment，and spent fuel pool purification. The treatment of radioactive waste resin must consider system radiation protection issues and adopt effective shielding measures to protect public and environmental safety. Design three different shielding combinations based on the type and location of shielding materials，calculate and analyze the shielding effects and applicable scenarios of each method，and provide relevant reference for engineering applications. Meanwhile，using SuperMC radiation shielding calculation software，the surface dose rate of waste resin flowing through equipment such as reactor vessels，resin grinding devices，waste resin transfer tanks，and booster pumps was calculated and analyzed with waste resin treatment facilities as the research object. The calculation results indicate that the surface dose rate of the relevant equipment is 22.13-48.64 μSv/h without shielding. The use of a lead shielding layer to shield and protect equipment with higher surface dose rates meets the requirements of a control area of 5 μSv/h or less after shielding.

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The study assessed the acceptability of the currently existing approach to calculating shielding in computed tomography rooms and formulated proposals for its improvement, taking into account the parameters of research in domestic medical practice. It is proposed to use the dose-length product as the main measured parameter in the new approach to calculating shielding. Typical tomograph workload values of dose-length product, typical for examinations in the Russian Federation, were assessed. Using thermoluminescent detectors, the experimental assessment of the distribution of absorbed dose in the air in a computed tomography room was carried out. The experimental results showed that the distribution of of X-ray exposure in the central plane is not isotropic. Due to the attenuation of exposure by the gantry design, “shadow” zones are observed, where the absorbed dose reduction in the air reaches 10-13 times compared to its level in the direction of couch movement. Based on the results of measurements of the distribution of absorbed dose in air, the relative scatter coefficients were calculated. The workload was determined for 7 tomographs (1 in St. Petersburg and 6 in Moscow). For each tomograph, the average weekly number of head and body examinations and the values of dose-length product for these areas were assessed, taking into account the number of multiphase examinations. The obtained workload values turned out to be one or two orders of magnitude higher than those currently used in calculating shielding. Based on the results of the study and literature data, the work proposes a new approach to calculating shielding in computed tomography rooms, based on the value of measured dose-length product during the examination and allowing to take into account the specifics of the operation of the device.

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.

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.)

BackgroundFusion reactors have the characteristics of high neutron energy, large spatial scales, complex geometric structures, and large radiation attenuation gradients, which make nuclear design and analysis more challenging.PurposeThis study aims to verify the accuracy of the calculation of Monte Carlo code cosRMC on the material shielding effect in fusion reactor applications.MethodsTwo deep penetration benchmarks, i.e., FNS-Duct (Fusion Neutron Source Dogleg Duct Streaming) and FNS-W (FNS Clean Experiment on Tungsten Cylindrical Assembly) from the Shielding Integral Benchmark Archive Database (SINBAD), were taken for simulation study of the penetration behaviors of neutrons in models under different geometric shapes and materials. First of all, detailed cosRMC models was established based on benchmark experimental models, neutron flux spectra of detectors at different penetration depths in benchmark experiments were calculated using cosRMC. Then, the variance reduction technique of cell importance was applied to improving the accuracy of cosRMC in calculating deep penetration problems. Finally, neutron flux spectra calculated by cosRMC were compared with those calculated by MCNP5, as well as the experimental results of the benchmarks.ResultsThe comparison results show that the calculations of cosRMC conform to that of MCNP5 with relative deviations of less than three times the standard errors. Both calculation results match well with the experimental values.ConclusionsThe study verifies that cosRMC can be applied to shielding calculations of fusion reactors.

The lack of a universal and effective Global Variance Reduction (GVR) method for the deep-penetration Monte Carlo calculation makes it difficult or impossible to calculate the radiation field of large nuclear facilities. We propose a Density-extrapolation Global Variance Reduction (DeGVR) method for the large-scale radiation field calculation. After reducing the density of all materials to avoid the deep penetration problem, two global flux distributions of the two different material densities are obtained with two fixed-source calculations, and then the global flux distribution of the original material density is obtained by extrapolating the two global flux distributions. The global flux distribution of the original material density is then used for global variance reduction. In a large-scale model with a flux attenuation of 1080, the DeGVR method only takes 40.3 minutes to build the radiation field, improves the Average Figure-of-Merit (AV.FOM) up to 85652 times and the counting rate per time (CRPT) up to 88107 times compared with the standard Monte Carlo method. In addition to the high efficiency, the DeGVR method has high universality and robustness because it constructs the global information in a clever way that does not rely on the complex mathematical derivation and geometric modeling. The DeGVR method shows excellent application potential in large-scale radiation analysis. • Scientific Importance: provides a clever solution to the deep penetration problem. • Effectiveness: improves the Average Figure-of-Merit (AV.FOM) greatly. • Robustness: without complex mathematical derivation and geometric modeling. • Engineering application: is helpful for large-scale radiation field calculation. [ABSTRACT FROM AUTHOR]

Multigroup constants are the foundation of neutron and photon transport problems, and the accuracy of multigroup cross-sections has a significant impact on shielding calculation. Challenges have arisen in generating accurate multigroup macroscopic cross-sections for some problems using the widely used cross-section processing code TRANSX 2.15. We developed the multigroup neutron-photon macroscopic cross-section processing module in the shielding transport code ARES. The module is capable of handling the neutron-photon coupled master libraries in MATXS format and providing the problem-dependent multigroup macroscopic cross-sections tailored to each specific shielding physics problem. The self-designed problems with a single nuclide, as well as the iron and OKTAVIAN experiments, are used to verify and analyze the accuracy of neutron and photon macroscopic cross-sections. Results indicate that the macroscopic cross-sections, neutron flux and neutron-photon leakage spectrum obtained by the MGMXS module are consistent with corresponding reference values. As for the JANUS Phase I fixed source shielding benchmark, the relative differences in the reaction rate between the calculation results and experimental data are within 20%. The module provides better problem-dependent multigroup macroscopic cross-sections for neutron and photon shielding calculations that satisfy the accuracy requirements of engineering applications. [ABSTRACT FROM AUTHOR]

The shielding calculation of neutron streaming problems with ducts is characterized by the strong anisotropy of angular flux, which poses a challenge for the analysis of nuclear installations. The discrete ordinate method is one of the most commonly deterministic techniques to solve the neutron transport equation, in which the accuracy and efficiency neutron are crucial to ensure the reliability of the streaming shielding simulation. We implemented the goal-oriented multi-collision source algorithm in the 3D transport code ARES. This algorithm can determine the importance factor based on the adjoint transport calculation, obtain the response function to enable problem-dependent, goal-oriented spatial decomposition, and provide the error estimation as a driving force behind the dynamic quadrature to optimize the source iteration. This study focuses on verifying the goal-oriented multi-collision source algorithm under the neutron streaming problems, and the capabilities of the algorithm have been tested on IRI-TUB benchmark of SINBAD database. The numerical results show that the algorithm can effectively control the angular discretization error for the neutron streaming problems, which is more economical than the traditional discrete ordinate calculation. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center 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.)

BackgroundThe numerical results from a neutron source, as the important input parameter for the transport calculation, directly affect the accuracy of shielding calculations for reactor. The apparent differences between core sources are related to their geometric model, burnup, and power distribution.PurposeThis study aims to improve the calculation accuracy of the core neutron source for nuclear reactor shielding.MethodsFirstly, the geometric weight of each component in the core was generated by analyzing the characteristics of the radial source distribution on the basis of neutron importance, and the fine source mesh calculation was conducted for the peripheral components with high geometric weight and the region with a large power gradient. Then, a layered approach for different axial height positions was employed to reduce the influence of the axial power peak factor to achieve stable transport calculation results, and the source and geometry meshes were mapped according to the volume weight method to ensure the conservation of total source. Finally, the NUREG/CR-6115 core as a benchmark model was used for numerical verification.ResultsNumerical verification results indicate that, compared with the average source calculation, the multi-weight source mesh mapping algorithm reduces the root mean square of the relative error in the fast neutron fluence by 18.46% between the transport calculation results and the reference value.ConclusionsThe multi-weight source mesh mapping algorithm can be employed to obtain an accurate source distribution, improve the accuracy of the shielding calculation, and satisfy the requirements of engineering applications.

The discrete ordinates method (SN) is one of the major shielding calculation method, which is suitable for solving deep-penetration transport problems. Our objective is to explore the available quadrature sets and to improve the accuracy in shielding problems involving strong anisotropy. The linear discontinuous finite-element (LDFE) quadrature sets based on the icosahedron (in short, ICLDFE quadrature sets) are developed by defining projected points on the surfaces of the icosahedron. Weights are then introduced in the integration of the discontinuous finite-element basis functions in the relevant angular regions. The multivariate secant method is used to optimize the discrete directions and their corresponding weights. The numerical integration of polynomials in the direction cosines and the Kobayashi benchmark are used to analyze and verify the properties of these new quadrature sets. Results show that the ICLDFE quadrature sets can exactly integrate the zero-order and first-order of the spherical harmonic functions over one-twentieth of the spherical surface. As for the Kobayashi benchmark problem, the maximum relative error between the fifth-order ICLDFE quadrature sets and references is only −0.55%. The ICLDFE quadrature sets provide better integration precision of the spherical harmonic functions in local discrete angle domains and higher accuracy for simple shielding problems.

Multigroup constants are the foundation of neutron and photon transport problems, and the accuracy of multigroup cross-sections has a significant impact on shielding calculation. Challenges have arisen in generating accurate multigroup macroscopic cross-sections for some problems using the widely used cross-section processing code TRANSX 2.15. We developed the multigroup neutron-photon macroscopic cross-section processing module in the shielding transport code ARES. The module is capable of handling the neutron-photon coupled master libraries in MATXS format and providing the problem-dependent multigroup macroscopic cross-sections tailored to each specific shielding physics problem. The self-designed problems with a single nuclide, as well as the iron and OKTAVIAN experiments, are used to verify and analyze the accuracy of neutron and photon macroscopic cross-sections. Results indicate that the macroscopic cross-sections, neutron flux and neutron-photon leakage spectrum obtained by the MGMXS module are consistent with corresponding reference values. As for the JANUS Phase I fixed source shielding benchmark, the relative differences in the reaction rate between the calculation results and experimental data are within 20%. The module provides better problem-dependent multigroup macroscopic cross-sections for neutron and photon shielding calculations that satisfy the accuracy requirements of engineering applications.

BackgroundThe high-fidelity neutron transport calculation requires refined geometric modeling whilst the unstructured meshes have strong adaptability to copy with the changes bring by complex geometry structure, and overcome the deficiencies of structured meshes in modeling capability.PurposeThis study aims to develop and validate a two-dimensional shielding calculation code ThorSNIPE which can be used to improve the modeling ability for analysis complex problems.MethodsFirst of all, problem solving model was established with discrete ordinates method and finite element method on the basis of the first order Boltzmann transport equation. The computational performance of continuous finite element method and discontinuous finite element method were compared and analyzed. Mass-matrix lumping technique was further applied to improve the reliability of solving model. Then, a two-dimensional discrete ordinate-finite element shielding calculation program ThorSNIPE was developed on the basis of above model. Finally, the code was validated by BWR cell critical benchmark, Argonne-5-A1 fixed source benchmark and Dog leg duct benchmark.Results & ConclusionsThe numerical results show that calculation value provided by ThorSNIPE is in good agreement with reference value, indicating that ThorSNIPE is suitable for complex shielding calculation, and Mass-matrix lumping technique can effectively suppress the non-physical spatial oscillations without reducing the calculation accuracy.

The shielding calculation of neutron streaming problems with ducts is characterized by the strong anisotropy of angular flux, which poses a challenge for the analysis of nuclear installations. The discrete ordinate method is one of the most commonly deterministic techniques to solve the neutron transport equation, in which the accuracy and efficiency neutron are crucial to ensure the reliability of the streaming shielding simulation. We implemented the goal-oriented multi-collision source algorithm in the 3D transport code ARES. This algorithm can determine the importance factor based on the adjoint transport calculation, obtain the response function to enable problem-dependent, goal-oriented spatial decomposition, and provide the error estimation as a driving force behind the dynamic quadrature to optimize the source iteration. This study focuses on verifying the goal-oriented multi-collision source algorithm under the neutron streaming problems, and the capabilities of the algorithm have been tested on IRI-TUB benchmark of SINBAD database. The numerical results show that the algorithm can effectively control the angular discretization error for the neutron streaming problems, which is more economical than the traditional discrete ordinate calculation.

The Particle and Heavy Ion Transport code System (PHITS) is a general-purpose Monte Carlo simulation code that has been applied in various areas of medical physics. These include application in different types of radiotherapy, shielding calculations, application to radiation biology, and research and development of medical tools. In this article, the useful features of PHITS are explained by referring to actual examples of various medical applications. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

• Radiation safety evaluation of KSTAR is conducted with operating conditions for duplicated neutron yields • Dose assessment is conducted for radiation workers and major work areas • An activation analysis of the main tokamak structures is carried out. • An air activation assessment was performed and compared with standards of effluent level. To determine the viability of the 2-fold neutron yield increase of the Korea Superconducting Tokamak Advanced Research (KSTAR) owing to the installation of a second neutral beam injection system to increase the plasma density and a divertor material change from graphite to tungsten to reduce the heat load, we performed a radiological evaluation and shielding analysis of an existing building structure and the KSTAR with an increased neutron yield per shot. According to the shielding analysis results, the weekly and annual doses outside the building did not exceed the criteria, and the fluxes and doses in the plasma experimental room were unaffected by the divertor material change. The maximum dose rate in the working area in the tokamak was 2.16E+03 μSv/h at the surface-plasma-facing components. However, the dose rate was reduced to the background level one month after a 1-shot operation of 300 s. Ar-41 was the major nuclide for the air activation calculations; however, the effluent level exceeded the derived air concentration after the 1-shot operation ended. The residual activity concentration of Ar-41 was reduced below the derived air concentration after 243 min. The annual operating condition of the KSTAR is suggested based on the evaluation results. [ABSTRACT FROM AUTHOR]

• A new automatic modeling of discrete ordinate program (AMDOP) is developed to automatic convert CAD model to 3D neutronics physics shielding calculation model for the discrete ordinate program. • The linear volume-weighted homogenization method is proposed to mix the material of the mesh containing two or more materials, and obtain the high fidelity neutronics physics shielding calculation model. • The AMDOP is verified based on the Cartesian mesh model and cylindrical mesh model, and applied to the NUREG-CR-6115 benchmark. The azimuthal distribution of scalar flux above 1 MeV are calculated at five different locations in the pressure vessel, and the calculation results show that the automatic modeling method can effectively improve calculation accuracy. In view of reactor shielding calculations with complicated geometry structure, traditional manual modeling of discrete ordinate program is a very cumbersome, inefficient, and error prone. CAD visualization modeling of complicated shielding geometry is implemented based on the open source platform for numerical simulation (SALOME). The discrete ordinate program often requires millions or tens of millions discrete meshes to build a neutronics physics model for a three-dimensional complexity model, each discrete mesh needs to be assigned material. Currently, center point material mapping method is commonly used to assign the mesh material, which assign the material at the center point of the mesh to the mesh material. For the complicated geometry, the discrete meshes typically have multiple materials, which introduces modeling errors. Linear volume-weighted homogenization method is proposed to obtain the uniform mix material of the mesh containing two or more materials. Automatic modeling of discrete ordinate program (AMDOP) is developed based on the linear volume-weighted homogenization method d, which can achieve precise automatic conversion of CAD models to 3D neutronics physics shielding calculation model for the discrete ordinate program. The AMDOP is verified based on a cylindrical model and Kobayashi model, the calculation results of automatic modeling are completely consistent with those of manual modeling, and the linear volume-weighted homogenization method can effectively improve calculation accuracy for model with multiple materials in the discrete meshes. To further verify the reliability and practicality of AMDOP, the practical engineering shielding calculation benchmark problem (NUREG-CR-6115) is modeled and calculated, the results calculated by the linear volume-weighted homogenization method are in good agreement with the reference results, this indicates that the AMDOP automatic modeling process is correct and reliable. [ABSTRACT FROM AUTHOR]

• Weight window technique improves radiation-shielding calculations. • Utilizing recursive Monte Carlo (RMC) method with TRAWEI code efficiently generates optimal weight parameters. • RMC is beneficial for deep penetration shielding calculations. 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. [ABSTRACT FROM AUTHOR]

BackgroundDeterministic method is widely used in shielding calculation for fusion and fission devices. The data accuracy of multi-group cross section has significant influence on the deterministic calculation results.PurposeThis study aims to obtain a multi-group cross-section library suitable for various shielding problems by developing a multi-group cross-section processing code ARES-MACXS based on Bondarenko resonance processing method with functionalities of resonance processing, temperature interpolation, transport correction, up-scattering processing, collapse group, and so on, and uses the exponential interpolation method to obtain more accurate multi-group cross sections.MethodFirst of all, KASHIL_E70 library and FENDLL-3.1d library were respectively processed by ARES-MACXS for single nuclide calculation to verify the correctness of the code, and the results were compared with continuous energy Monte Carlo code. Then, the ill-Fe, OKTAVIAN-Al and other benchmarks were calculated to verify the applicability of the code to different nuclides, and the results were compared with the continuous energy Monte Carlo code. Finally, the ill-Fe, OKTAVIAN-Al and other benchmarks calculated by ARES-MACXS were compared with experimental measurements.ResultsThe error between the deterministic transport calculation result and the Monte Carlo calculation result is less than 1%, the error between the deterministic transport calculation result and experimental measurement values is less than 20%.ConclusionsThe results show that the multi-group cross sections calculated by ARES-MACXS have reliable accuracy and can be applied to the verification of benchmarks, hence can be applied to the calculation of engineering problems.

The components from spent fuel are discharged from the core and then stored in the framework of the spent fuel pool for cooling. However, it is of great significance to save the storage space of the spent fuel pool by shrinkage and shearing work to increase the spare fuel lattice number. In order to solve the problem of shrinkage and shearing work of spent fuel involving the problem of radioactive safety, the radioactive source item is calculated by ORIGEN2 program base on Unit 1 II of Lingao Nuclear Power Plant (NPP), and the radiation dose rate of the related component shrinkage operation scene is simulated by the MCNP5 program. In addition, the effectiveness of shielding measures is discussed, and the maximum dose rate is within 0.35 μSv / h at the distance of 2.5 m from component center, and the maximum dose rate is almost 0 at the distance of 3.2 m from the component center. The intensity of the radiation dose produced by the related components is very low and can be neglected, which belong to the green area of NPP. The program calculation system from source term calculation to shielding calculation is established, and an engineering example is referenced, and its application and analysis are carried out. It provides a basis for radioactive safety analysis and evaluation for the shrinkage operation of spent fuel and makes the shrinkage technology of fuel-related components safer and more reliable. [ABSTRACT FROM AUTHOR]

Highlights • A regional angular adaptive algorithm to reduce angular discretization errors for shielding calculations. • The accuracy of this algorithm with FENDL/MG-3.0 libraries was investigated through analyses of streaming experiments with deuterium–tritium neutron sources. • The ARES results show good agreement with experimental values, for this narrow and deep gap structure, the calculation accuracy for the streaming analyses is mostly within ± 25% for fast neutron reaction rates. Abstract Shielding analyses for fusion reactors pose significant computational challenges, including strong anisotropic scattering and a combination of deep-penetration shielding and complex channeling. The discrete ordinates method is one of the most widely used techniques to solve the linear Boltzmann equation. We propose a regional angular adaptive algorithm to reduce angular discretization errors for shielding calculations. Angular adaptive refinement techniques are based on estimated local errors. The local errors are weighted by the importance distribution toward the computational goal, providing an appropriate goal-oriented adaptivity. The accuracy of this algorithm with FENDL/MG-3.0 libraries was investigated through analyses of streaming experiments with deuterium–tritium neutron sources at JAERI/FNS. From comparing reaction rates of measurements and calculations, the angular adaptive algorithm suffices in predicting the fast neutron field inside the gap. The solutions show that the algorithm can improve the accuracy of calculation results. [ABSTRACT FROM AUTHOR]

For large-scale pressurized water reactor (PWR) deep-penetration shielding calculation, Monte Carlo (MC) method requires simulating a great number of particles to obtain reliable results, thus huge amount of computation time is the main problem of the MC method. Discrete Ordinates-Monte Carlo (SN-MC) coupling method combines the advantage of the SN method with high efficiency and the MC method with fine geometrical modeling. This paper studies the 3D SN-MC coupling method, generates surface source using the angular fluence rates of a SN code JSNT, and develops a source sampling subroutine for a MC code JMCT. The coupling method is verified on measurements and applied to CAP1400 cavity streaming calculation. Numerical results show that the coupling method increases calculation efficiency by 2–10 times for deep-penetration shielding calculation with high precision compared with reference results. The coupling method is effective for large-scale PWR deep-penetration shielding calculation. [ABSTRACT FROM AUTHOR]

With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be run with TRIPOLI-4 for the fusion neutronics study. This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. [ABSTRACT FROM AUTHOR]

The variance reduction technique of mesh-based weight window was widely used in reactor shielding calculation. However, there are disadvantages in the traditional calculation procedure with weight window method based manual operation. It is time consuming to perform iterative calculations to get an effective weight window. Even for experts, they can only give an approximate division range of mesh parameters. Iteratively mesh plotting is needed for the under-bias or over-bias of particles. To solve the problem, an automatic adaptive mesh generation method was proposed for weight window based on the characteristics of particle’s transport. The method involved coarse and fine meshing. Coarse mesh boundary parameters can be automatically generated based on the division of the interested area. Fine mesh boundary parameters can be produced by the automatic process of the mean free paths at the current energy of the particle. Coarse meshing and fine meshing were combined to generate the final meshing scheme. The fusion shielding benchmark and the ITER T426 benchmark were selected as the benchmark cases. The accuracy and efficiency of the proposed method were demonstrated. [ABSTRACT FROM AUTHOR]

Concretes are popularly used as shielding materials in nuclear radiation facilities. Special concrete features are employed in the facilities where Fast neutrons, including corresponding 'prompt gamma' rays, are prevalent. The conventional shielding calculations for fast neutrons are complicated and constrained with energy-limited formulas. In this research work, a concrete bunker model has been developed with a variety of compositions and densities. Eleven following appropriate composite materials were used in this case; G4_C (2.3), Ordinary (2.5), Barytes BA-a (3.5), Magnetite M-a (3.55), Ilmenite I-Ia (3.5), Limonite with Steel Punch LS-a (4.54), (80%G4_C+20%Fe+10%B) Mix-1 (2.9), (80%G4_C+20%B4C) Mix-2 (2.35), (80%G4_C+20%BaSO4) Mix-3 (2.74), Serpentine (3) and Serpentine (3.3). Depth-dose profiles for each concrete composition were investigated by Monte Carlo simulation, subjected against an 18.5 MeV neutron source. Shielding properties, like Relaxation lengths, Half Value Layers, and Tenth Value Layers have been deduced harnessing these data. The most efficient materials for shielding fast neutrons regarding half value layers found for the two physics lists, QGSP_BERT_HP and Shielding, successively are, 5.78 and 12.48 cm for Serpentine (3.3), 14.88 and 9.18 cm for Ilmenite I-Ia (3.5), 14.28 and 25.18 cm for Limonite with Steel Punch LS-a (4.54), and 13.58 and 13.48 cm for (80%G4_C+20%Fe+10%B) Mix-1 (2.9). This model could be used to determine the radiation shielding parameters made of any kind of materials and composites using different sources of radiation. • Development of a bunker model using Monte Carlo simulation tool Geant4. • Shielding properties of 11 concrete types against a 18.5 MeV fast neutron source. • Dose depth profile analysis, HVL, TVL and RL for each sample material. • Serpentine, Ilmenite, Limonite and Fe–B mixture most effective for fast neutrons. • A flexible model to investigate radiation shielding properties of any materials. [ABSTRACT FROM AUTHOR]

In the enclosure of synchrotron radiation experiments using a monochromatic beam, secondary radiation arises from two effects, namely fluorescence and scattering. While fluorescence can be regarded as isotropic, the angular dependence of Compton scattering has to be taken into account if the shielding shall not become unreasonably thick. The scope of this paper is to clarify how the different factors starting from the spectral properties of the source and the attenuation coefficient of the shielding, over the spectral and angular distribution of the scattered radiation and the geometry of the experiment influence the thickness of lead required to keep the dose rate outside the enclosure below the desired threshold. [ABSTRACT FROM AUTHOR]

A simple estimate for the thickness of the shielding of photon beam lines with energies below the pair production threshold has been developed. In this energy range scattering and fluorescence are the main sources of secondary radiation. In the low energy regime up to about 150 keV fluorescence dominates. This justifies the assumption of an isotropic radiation field allowing the derivation of a simple expression in which the influence of the various parameters like energy bandwidth or distance from the primary beam become clearly visible. Furthermore the simplicity of this expression allows the generation of a spread shield calculation resulting in a simple estimate of shielding thicknesses. [ABSTRACT FROM AUTHOR]

Abstract: The Monte Carlo (MC)-discrete ordinates (SN) coupled method is an efficient approach to solve shielding calculations of large nuclear device with complex geometries. However, it is time-consuming and error-prone to describe and verify the models for simulation. The MC-SN coupled automatic modeling program for radiation transport simulation (RCAM) has been developed to achieve bi-directional conversion between CAD model and input geometry description of MC-SN coupled transport simulation code. This paper introduces main features of RCAM. RCAM-based modeling process and preliminary applications for ITER benchmark model are also presented. [Copyright &y& Elsevier]

Radiation shielding calculations were performed for the shielding design around beam injection and beam extraction areas at the 3-GeV proton synchrotron in the Japan Proton Accelerator Research Complex (J-PARC) with the PHITS Monte Carlo code. Prompt dose rates outside the shields were obtained from calculated neutron energy spectra and flux-to-dose conversion coefficients prepared for the shielding design in J-PARC. The thicknesses of shields were determined from the comparison results of the prompt dose rates with the criteria of dose limit for the design in J-PARC. [ABSTRACT FROM AUTHOR]

The neutron streaming through slits between the shielding blanket modules in DT fusion reactors are evaluated, and the helium production in the vacuum vessel, the nuclear heating density and the radiation damage in the super-conductive magnet are calculated by using three-dimensional Monte Carlo method. By changing systematically the blanket and the vacuum vessel thicknesses, the boron content in the vacuum vessel, and the slit width, the analytical representations on the helium production, the nuclear heating density and the radiation damage along the slit are established as functions of the slit width, thickness in the blanket and the vacuum vessel, and the boron content based on the Monte Carlo results. Their dependencies on the blanket and the vacuum vessel compositions are also clarified. The analytical formula thus-obtained is applied to clarify the shielding design conditions required to satisfy the shielding design criteria for DT fusion reactor. [Copyright &y& Elsevier]

To solve the Boltzmann transport equation, the discrete ordinates method was applied using multi-group cross-sections, the accuracy of which affects markedly the transport calculation. For large shielding models, using fine-group cross-sections directly in the transport calculation is inefficient. Subsequently, the multi-group cross-section generation code ARES-MACXS utilized contribution theory in selecting a broad-group structure to meet calculational requirements of engineering shielding. For different computational problems, the ARES-MACXS code generates different optimal broad-group structures and corresponding multi-group working cross-sections. The numerical results indicate that, compared with the BUGLE-B7 broad-group structure, using these optimal broad-group structures may yield transport calculation results of higher accuracy with fewer groups. On the basis of a comprehensive analysis of different shielding calculation problems, ARES-60 broad-group structure with 46 groups of neutrons, 14 groups of photons is proposed, which is suitable for AP1000 shielding calculations and other similar PWR shielding calculations. • Multi-group cross-section generation code for radiation shielding calculation. • Optimization algorithms based on contribution theory for broad-group energy structures(ESs). • Broad-group ESs for fast/thermal flux, detector response and dose rate. [ABSTRACT FROM AUTHOR]

• Using neural network for the reconstruction of radiation fields. • Multi-sampling method to reduce flux/dose gradient fluctuation and improve reconstruction accuracy. • The reconstruction performance of the proposed neural network is better than that of the conventional neural network. In order to achieve optimal radiation protection, rapid and accurate reconstruction of radiation field has vital significance in the selection of working paths during the overhaul of nuclear power plants and the decommissioning of nuclear facilities. The radiation field is usually reconstructed by various interpolation methods, but the reconstruction accuracy of such methods is insufficient, With the improvement of AI technology, neural networks have great potential in radiation field reconstruction, but conventional neural networks is prone to local minima and vanishing grandient problem. This paper aims to develop a radiation field reconstruction method based on an adaptive Back-propagation (BP) neural network neural network method with learning rate decay and a corresponding sampling method for multisampling in places where flux gradient changes drastically, and verify its accuracy and feasibility. The proposed method achieves global optimality and avoids vanishing grandient problem by virtue of adaptive algorithm and learning rate decay, ensuring that the radiation field is reconstructed with the smallest relative average error when the sampling point is determined, moreover, the proposed sampling method can greatly improve the accuracy of radiation field reconstruction. The accuracy of the proposed method was tested with three MC simulated radiation fields with simpler cases, and the feasibility of the proposed method was further validated with two MC simulated, more complex and realistic scenes. The results of the proposed method show that the errors of the three test cases are 1.7%, 6.8%, and 7.8%, and the errors of the two validated cases are 8.8% and 7.7%, respectively. The merit of this method was preliminarily verified, further validation is underway to validate its application in real world scenarios. [ABSTRACT FROM AUTHOR]

• A parallel multicollision source (MCS) method to solve discrete ordinates equation. • The MCS method calculates multicollision sources via spatial decomposition. • The MCS method can eliminate secondary ray effects and numerical diffusion. Discretization errors like ray effects and numerical diffusion are major shortcomings of S N calculation in shielding problems involving strong anisotropy. Especially in isolated point source shielding problems with cavities and ducts, the angular flux distribution in spatial-angular space may become unsmooth and discontinuous because of heterogeneity effects in the first few iterations. We have developed the multicollision sources (MCS) algorithm via spatial decomposition. The method semi-analytically calculates the scattering source for particles that have suffered several collisions in a selected region, and the S N transport sweeping calculates the scattering for particles in the remaining regions. The calculation parameters of the MCS method is selected according to the distribution of flux in different regions and the scattering effect of shielding materials. For practical complex multi-group problems, the MCS method can be carried out independently in several groups considering the specific problems. We implemented the MCS method in the 3D, parallel, multigroup discrete ordinates code ARES, and numerical results and evaluations are given for a self-designed problem and several general benchmarks. All these results can verify the ability of the MCS method to reduce discretization errors and demonstrate that the MCS method can improve calculation accuracy obviously with an acceptable additional computation cost. [ABSTRACT FROM AUTHOR]

In this study, experimental measurements were performed on the spectra of neutrons which penetrate concrete and steel of various thicknesses values when a proton beam of 24 GeV/c was incident on a copper target at the CHARM facility in the East Hall of the CERN Proton Synchrotron (PS) The thicknesses of concrete and steel ranged up to 360 cm and 80 cm, respectively. To measure the neutron spectra, an NE213 scintillator was positioned on the top roof of the shielding structure as the neutron detector. The light output distributions of the detector were converted into the neutron energy spectra using the unfolding method with a calculated response matrix after removing the γ -ray and charged particle events by pulse-shape discrimination and veto counter signals, respectively. The neutron spectra were in agreement with the results obtained using the Monte Carlo simulation code, PHITS, within a factor of 1.4 except for the case of steel 80 cm. The attenuation profiles for concrete and steel were consistent with previous foil activation results within the respective uncertainties. [ABSTRACT FROM AUTHOR]