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In this note, we prove that for two compatible fuzzy metrics M H and M K on H and K, respectively, there exists a fuzzy metric M on H ⋃ K such that M | H = M H and M | K = M K under the conditions that t-norm ∗ is positive and fuzzy metrics M H , M K are strong, or t-norm ∗ is positive and t-norm ∗, fuzzy metrics M H , M K satisfy the Lipschitz conditions. [ABSTRACT FROM AUTHOR]

This technical note presents a unified framework of stabilized finite element methods for solving the Boltzmann transport equation. The unified framework is derived from the standard Galerkin weak form with a subgrid scale model, which is different from the traditional Petrov-Galerkin finite element framework that modifies the test function to construct the stabilization term. By this method, first, the unknowns are decomposed into their numerical solutions and residuals. The decomposed unknowns are then embedded into the Galerkin weak form with an approximation for the residual, which yields a stabilized variational formula. Different methods of stabilization are derived from different approximations of the residual. Under this framework, all the frequently used stabilized methods can be obtained, including the streamline upwinding Petrov-Galerkin method, the Galerkin least-squares method, and the algebraic subgrid scale method. Thus, a unified framework of such methods is established. The similarities and differences across the different approximations are also compared in this technical note. The numerical results show that the behaviors of different methods are similar with the same stabilization parameters and that all these stabilized techniques can yield satisfactory and stable solutions. [ABSTRACT FROM AUTHOR]

The unstructured mesh (UM)–based Monte Carlo (MC) method can utilize modern computer-aided-design/computer-aided-engineering platforms to obtain geometric models with reduced human effort and is capable of generating high-resolution tally data. This approach presents a significant advantage over the traditional Constructive Solid Geometry (CSG)–based MC method in handling complex geometries and conducting multiphysics calculations. In this study, the UM-based MC calculation capability was developed in the MC code NECP-MCX. On this basis, an automatic UM-based Consistent Adjoint-Driven Importance Sampling (CADIS) method was further studied and implemented in which the adjoint deterministic calculation, forward MC calculation, and variance reduction (VR) parameter generation were performed on the unified UM model. To achieve this, the discrete ordinates (SN)–Discontinuous Finite Element Method (DFEM) code NECP-SUN was embedded into NECP-MCX as the adjoint transport solver. Validations of the developed code and evaluations of the VR performance of the UM-based CADIS method were conducted on the Pool Critical Assembly (PCA) Replica benchmark and H. B. Robinson Unit 2 (HBR-2) benchmark. The numerical results indicated that the developed UM-based particle tracking capability achieved comparable accuracy to the CSG-based approach. Furthermore, compared to the traditional CADIS method, the UM-based CADIS method demonstrated higher figure-of-merit (FOM) values (3.5 to 44 times higher for the PCA Replica benchmark and 2.22 to 2.92 times higher for the HBR-2 benchmark), highlighting the superior VR performance of the UM-based CADIS method over the traditional CADIS method. [ABSTRACT FROM AUTHOR]

• Detailed analysis of different approximations in multi-group library processing. • Detailed analysis of different approximations in self-shielding calculation. • Providing reference for improving conventional methods and legacy codes. Approximations and treatments commonly applied in multi-group library and self-shielding calculation of conventional neutronics calculation codes are analyzed based on a numerical nuclear reactor physics code NECP-X. The numerical results show that the approximations and treatments introducing largest errors for UO2 pin cell problems are narrow resonance (NR) approximation, ignorance of multi-group (MG) equivalence effect, Bondarenko iteration method and using average fission spectra. For MOX pin cell problems, the approximations and treatments introducing largest errors are ignorance of thermal resonance, NR approximation, resonance interference factor method, ignorance of MG equivalence effect and in-flow transport correction. These analysis provide references and help to improve the conventional methods and legacy codes. [ABSTRACT FROM AUTHOR]

• The preliminary design of helium-cooled mixed bed breeder (HCMB) blanket for CFETR is performed. • The neutronics and thermal-hydraulics analyses of CFETR are performed. • The global TBR of HCMB is larger than that of HCCB. Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li 4 SiO 4 /Be 12 Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied. [ABSTRACT FROM AUTHOR]

The pseudo-resonant-nuclide subgroup method is a newly emerged self-shielding calculation method to treat resonance interference effect. It has been implemented in the NECP-X code which is developed for the high-fidelity neutronics calculation. Though this method has high accuracy, its computation time occupies more than half of the total computation time (mainly including resonance self-shielding calculation, multi-group transport calculation and depletion calculation). To improve the efficiency of the resonance self-shielding calculation, some analysis and optimization are performed in several aspects: firstly, the maximum number of subgroups is reduced from 7 to 5; secondly, the number of dilution cross section points is reduced from 21 to 9; thirdly, the number of fuel meshes in solving SFSE is reduced from 15 to 5; finally, number of fuel meshes in solving one-group fixed-source equation for the SPH factor is reduced from 15 to 1. The acceleration ratio of the self-shielding calculation is 3.45 for assembly without poison and 3.49 for assembly with poison. The time proportion of self-shielding calculation is reduced to less than 30%. The accuracy loss is negligible in the accelerations. Besides, it is found that the subgroup condensation scheme can obtain better self-shielded cross sections of component resonant nuclides than the dilution interpolation scheme. [ABSTRACT FROM AUTHOR]

The subgroup methods based on partial cross section fit scheme (PXSFS) and simplified partial cross section fit scheme (SPXSFS) are proposed in this paper to treat problems with non-uniform temperature distribution. These methods fit the cross sections at different temperatures as partial cross sections and share a same set of subgroup probabilities. The new methods are compared to the pre-existing methods: conventional subgroup method (CSM), the correlation model (CM), the subgroup level adjustment scheme (SLAS) and the number density adjustment scheme (NDAS). The numerical results show that the new methods can better predict the spatially dependent reaction rates than pre-existing methods. Within the new methods, the simplified scheme consumes less computation time and is more numerically stable. Additionally, the superhomogenization (SPH) correction method is studied, which is used to treat the multi-group (MG) equivalence effect. It is found that the subgroup-one-group (subgroup-1G) calculation can fully capture the MG equivalence effect. [ABSTRACT FROM AUTHOR]

• Development of the depletion/activation functionalities in NECP-MCX. • Coupling analysis of particle-transport/depletion/activation/source-term for CFETR. • The activation of Be should not be ignored for the depletion analysis of CFETR. As a bridge between ITER and Chinese DEMO, the design of Chinese Fusion Engineering Test Reactor (CFETR) is being carried out. The tritium breeding ratio (TBR) and the ShutDown-Dose-Rate (SDDR) are two significant neutronics parameters for CFETR and other fusion reactors. However, in the previous researches, the variation of the TBR with time was analyzed without consideration of the activation of beryllium. In the analysis of the SDDR, conventional methods assumes that the neutron spectra are not changed in the whole operation duration. To analyze the TBR and the SDDR with higher precision, the functionalities of the particle-transport/depletion/activation/source-term coupling analysis have been developed based on NECP-MCX, which is a Monte-Carlo-Deterministic particle-transport code developed by the Nuclear Engineering Computational Physics (NECP) Lab at Xi'an Jiaotong University. The transmutation solver NECP-Erica, developed also by the NECP Lab, is coupled with NECP-MCX internally. The integrated analysis code, which inherits the name NECP-MCX, is verified by several fusion-related experimental benchmarks. The verification results show the reliability of NECP-MCX. After the verification, the particle-transport/depletion/activation/source-term coupling analysis of CFETR is performed with NECP-MCX. Numerical results show that ignorance of the activation of beryllium overestimates the TBR of CFETR by 1.42% after an operation time of 16 years with a capability factor of 0.5. In the analysis of the SDDR, the ignorance of spectrum variation brings ∼2.0% discrepancy in the activity and decay heat of the structure materials. [ABSTRACT FROM AUTHOR]

The subgroup method is improved in several aspects to address challenges brought on by design features of the Fluoride salt-cooled High-temperature Reactor (FHR). Firstly, the Dancoff correction is applied to resolve the double heterogeneity arising from embedding TRISO fuel particles in the matrix of pebbles. Secondly, a fast Resonance Interference Factor (RIF) scheme is proposed to treat the resonance interference effect in the FHR. In this scheme, the heterogeneous system is converted into a homogeneous one according to self-shielded cross section conservation of the dominant resonant nuclide. The resonance interference effect is considered in the equivalent homogenous system by correcting the non-interfered self-shielded cross sections with RIFs which are obtained by solving the slowing down equation in hyper-fine energy group (∼1M number of energy groups). Finally, the resonance elastic scattering effect becomes considerable due to high temperatures in the FHR. This effect is considered by substitution of the conventional Resonance Integral (RI) table with that generated by the Monte Carlo method. The Monte Carlo method is modified via the Doppler Broadening Correction Rejection (DBRC) method to implement the Doppler broadened scattering kernel. The numerical results show that the Dancoff correction can significantly reduce errors brought about by the double heterogeneity. The fast RIF scheme provides more accurate effective self-shielded cross sections than the conventional iteration scheme. In addition, the speedup ratio of the fast RIF scheme is ∼3.3 compared with the conventional on-the-fly RIF schemes for TRU TRISO. The scheme to generate RI table can resolve the resonance elastic scattering effect encountered by the conventional scheme. [ABSTRACT FROM AUTHOR]

• A one-dimensional model for computing partial current discontinuity factor (PCDF) in subgroup method; • The multi-group equivalence effect can be captured in an acceptable precision by the PCDF based on 1D model; • There is negligible increase in computational time brought by the model. In the previous study, a non-iterative equivalence method is proposed, which guarantees that the reaction rates calculated by subgroup method in self-shielding calculation are preserved in subgroup-collapsed one-group calculation. This method shows advantage in efficiency over the Super-Homogenization (SPH) method. However, one-group fixed-source calculations are still needed for obtaining multi-group (MG) equivalence parameters. In this paper, a one-dimensional model is employed for obtaining the MG equivalence parameters in subgroup method. Based on the one-dimensional model, the one-group fixed-source calculation is simplified to one-angle tracking sweeping, which is time-saving. The numerical results show that the reaction rates can be preserved with the MG equivalence parameters obtained based on the one-dimensional model for pin cell problems. For fuel assembly problems, the k inf is improved by 140 pcm ∼ 187 pcm with negligible increase in computational time. [ABSTRACT FROM AUTHOR]

To determine reliable state parameters which could be used as early warning indicators of process failure due to the acidification of anaerobic digestion of food waste, three mesophilic anaerobic digesters of food waste with different operation conditions were investigated. Such parameters as gas production, methane content, pH, concentrations of volatile fatty acid (VFA), alkalinity and their combined indicators were evaluated. Results revealed that operation conditions significantly affect the responses of parameters and thus the optimal early warning indicators of each reactor differ from each other. None of the single indicators was universally valid for all the systems. The universally valid indicators should combine several parameters to supply complementary information. A combination of total VFA, the ratio of VFA to total alkalinity (VFA/TA) and the ratio of bicarbonate alkalinity to total alkalinity (BA/TA) can reflect the metabolism of the digesting system and realize rapid and effective early warning. [ABSTRACT FROM AUTHOR]

• A new hybrid Monte-Carlo-Deterministic particle-transport code, NECP-MCX, has been developed. • Development of NECP-MCX is focused on efficiency and high performance. • The figure of merit of NECP-MCX is ~103 higher than that of MCNP with weight window generator. A new particle-transport code NECP-MCX has been developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University. NECP-MCX is aimed at the simulation of deep-penetration problems, including reactor radiation-shielding calculation, fusion reactor blanket calculation, etc. These problems are challenging for the Monte Carlo (MC) method, which needs a huge number of particles to obtain reliable results. For the problems mentioned above, it is also challenging for the deterministic method to obtain results with high precision due to discretization errors. To overcome these challenges, NECP-MCX has been developed based on a hybrid Monte-Carlo-Deterministic method from scratch. The hybrid Monte-Carlo-Deterministic method utilizes the deterministic method to generate consistent mesh-based weight-window and source-biasing parameters for the MC method to reduce variance. The numerical results demonstrate that NECP-MCX is able to simulate deep-penetration problems with higher efficiency compared to the conventional MC codes. [ABSTRACT FROM AUTHOR]

• Two extension schemes of subgroup method are proposed to treat double heterogeneity. • Both of the schemes can treat double heterogeneity. • The subgroup XS correction scheme can be easily implemented in subgroup method. Two extension schemes based on the subgroup method are proposed to treat the double heterogeneity (DH) of fully ceramic micro-encapsulated (FCM) fuels. The first one is the hyper-fine energy group cross section (XS) correction scheme (HFCS), which corrects the hyper-fine energy group XSs with hyper-fine energy group disadvantage factors (DFs) before generation of resonance XS table and physical probability table. The second one is the subgroup XS correction scheme (SGCS), which corrects the subgroup XSs with subgroup DFs calculated by the equivalent homogenization method. These two scheme are implemented in the high-fidelity neutronics code NECP-X. A series cases are tested and the numerical results show that both of the HFCS and the SGCS can treat the DH and the precision of the HFCS is higher than that of the SGCS. [ABSTRACT FROM AUTHOR]

The pseudo-resonant-nuclide subgroup method (PRNSM) based global-local self-shielding calculation scheme was recently developed to resolve the spatial self-shielding and resonance interference effects for large-scale problems. But the PRNSM is not able to treat non-uniform temperature distribution. The reason is that the PRNSM generates physical probability tables at different temperatures with inconsistent subgroup probabilities. To overcome this defect, the on-the-fly subgroup method (OSM), which is an improvement of the PRNSM, is proposed. The new method generates the physical probability tables for different temperatures with shared subgroup probabilities and is able to treat non-uniform temperature distribution. This method replaces the PRNSM in the global-local self-shielding calculation scheme and is implemented in the high-fidelity neutronics code NECP-X. The numerical results show that the OSM can treat the spatial self-shielding effect, the resonance interference effect and the non-uniform temperature effect simultaneously. Besides, NECP-X is able to predict fuel temperature coefficients with high accuracy. • The on-the-fly subgroup method is proposed. • The new method generates physical probability tables for different temperatures with shared subgroup probabilities. • The new method can treat spatial self-shielding, resonance interference and temperature distribution effects. [ABSTRACT FROM AUTHOR]

Environmental pollution caused by organic pollutants has been attracted more and more attention by the researchers. In this paper, ZnIn2S4 modified TiO2 nanotube array (ZnIn2S4/TiO2) composite electrodes are prepared by loading chalcogenide semiconductor material ZnIn2S4 on the surface of TiO2 nanotube array (TiO2 NTA) for the degradation of methylene blue (MB). The physicochemical properties of ZnIn2S4/TiO2 composite electrodes are analyzed by X-ray photoelectron spectroscopy, X-ray diffraction, photoluminescence spectroscopy and so on. The photoelectrocatalytic degradation rate of ZnIn2S4/TiO2 composite electrode with the concentration of ZnIn2S4 as 0.5 mmol/L (ZISTO-2) reaches 93.3% under the bias voltage of 0.7 V and simulated sunlight irradiation for 180 min against MB, which is 22.44% higher than that of pure TiO2 NTA electrode, and the degradation rate of ZISTO-2 composite electrode toward MB is maintained at 78.2% after five cycles. The radical trapping experiments indicate that h+, ·O2− and ·OH has significant contribution in the photoelectrocatalytic degradation. This paper provides a new ideal way to solve the problem of water pollution under visible light. [ABSTRACT FROM AUTHOR]

Highlights • A multi-group effective cross section calculation method for FCM fuels is proposed. • The proposed method can exactly treat the self-shielding in TRISO particles. • The proposed method is capable of treating the FCM fuel consisting of multiple TRISO particles. • The proposed method is capable of treating the multi-layers of the TRISO coatings. Abstract A multi-group effective cross section calculation method for Fully Ceramic Micro-encapsulated (FCM) fuels containing stochastically dispersed tri-structural isotropic (TRISO) coated fuel particles is proposed to solve the double heterogeneity (DH). In resonance-energy range, the disadvantage factors are obtained by solving a one-dimensional model containing a TRISO particle with hyperfine group method. The matrix and TRISO particles will be homogenized by correcting the hyperfine-group cross sections with disadvantage factors. Due to the large absorption cross section of heavy isotopes in thermal-energy range, the spatial self-shielding effect in the TRISO particles should also be taken account. In the thermal-energy range, the multi-group disadvantage factors are obtained by the neutron's first-collision probabilities and penetrating probability equivalent. Based on the methods described above, the materials in the fuel rod are merged into a homogeneous material. The FCM fuel can be treated as traditional PWR lattice. In the present paper, the Dancoff correction factor of every rod is firstly obtained with neutron current method. Then a one-dimensional model for every fuel rod will be established by Dancoff factor equivalent. Finally, the hyperfine group calculation is carried out based on the one-dimensional rod model to get the effective cross sections of each fuel rod. Numerical results show that the proposed method is proved effective to treat DH for FCM fuel and capable of providing accurate effective cross sections. [ABSTRACT FROM AUTHOR]

The Iterated Fission Probability (IFP) method and Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Track-length importance Characterization (CLUTCH) method are commonly implemented in Monte-Carlo codes to calculate the sensitivity coefficients of k eff to continuous-energy nuclear data. However, these methods are plagued by the problem of large variance of sensitivity coefficients of the scattering reactions with relatively small cross sections. This problem means that more particles are needed to accurately calculate these sensitivity coefficients. To address this problem, a novel technique called CLUTCH coupled with forced collisions (CLUTCH-FC) method is proposed to reduce the variance of sensitivity coefficients of those reactions. The forced collision method is performed with specified reactions rather than specified cells in tradition to increase the sampling of specified reactions and a new rule for the setting of cutoff weight of neutrons is also proposed to make the best use of forced collisions. Three methods, viz. IFP, CLUTCH and CLUTCH-FC methods, are implemented in the Monte-Carlo code NECP-MCX. The verification is conducted in Godiva, Jezebel and TMI-1 problems by comparing the sensitivity coefficients calculated by IFP and CLUTCH methods with those calculated by the direct numerical perturbation (DNP) method. The numerical results indicate that 1) the sensitivity coefficients calculated by the IFP and CLUTCH methods corroborate well with those calculated by the DNP method, and 2) the CLUTCH-FC method is efficient in improving the tally efficiency of sensitivity coefficients of scattering reactions with relatively small cross sections by comparing results of CLUTCH-FC and CLUTCH methods. [ABSTRACT FROM AUTHOR]

• Random Vector Method for modelling dispersed TRISO particles in the fuel pebble has been introduced. • Vertically stacked hexagonal lattices of randomly distributed fuel and graphite spheres have been used for core filling. • Volume map has been introduced for the precise accounting of fuel and graphite balls. • The methodology has been automated for swift modeling of PBRs core with varying parameters e.g., core ht, F:M ratio, and the CPs. • Various analyses have been carried out and compared with the benchmark results in IAEA TECDOC 1382 and IAEA TECDOC 1694. • A fair agreement between the results and benchmark data highlights the suitability of this modeling approach for neutronic analysis in PBRs. Pebble bed reactors are particularly characterized by stochastic distribution of TRISO (Tri-structural Isotropic) particles and pebbles in the core. These random distributions pose a unique challenge for neutronic analysis. Generation of randomized locations for TRISO particles in fuel sphere along with pebbles in the core manually is a laborious and protracted task. This paper introduces an innovative and simple approach to model randomly distributed TRISO particles within fuel pebbles i.e., Random Vector Method. A three-dimensional grid is superimposed on fuel region of the pebble. A vector with random magnitude and direction starts from the center of grid. A TRISO particle is added if the mesh at the current position is empty. The vector is resampled if mesh is already filled or goes beyond the fuel region boundary. This repeats until required fuel loading has been achieved. In effect some of the mesh points remain empty, thereby imparting randomness. Secondly, each TRISO particle is arbitrarily displaced from the center within the margins of its cube. This acts as a twofold process resulting in random dispersion of TRISO particles in the fuel sphere. As for the pebbles, technique described in IAEA benchmark document has been applied. Fuel and moderator pebbles are stochastically distributed on a layer of hexagonal lattice while maintaining a certain fuel-to-moderator ratio. These layers are stacked vertically to achieve the desired core height. Volume map is introduced and applied to correctly account for cut spheres in the total tally of fuel/moderator balls for a given core configuration. Overall automation of the methodology proves to be highly useful in modelling pebble bed reactors with varying parameters e.g., core height, fuel to moderator (F:M) ratio, and number of TRISO particles. It also significantly reduces the time and effort required to generate these complex random distributions, making it a useful tool for various pebble bed reactor calculations. To validate the methodology, the HTR-10 reactor has been modelled using this technique, and the results are compared with various benchmarks from the IAEA TECDOC 1382 and 1694. The findings demonstrate a good concordance with the benchmark data, highlighting the suitability and efficiency of this modelling technique for neutronic analysis in pebble bed reactors. [ABSTRACT FROM AUTHOR]

• A new Dancoff based equivalent model is proposed. • Multi-group disadvantage factor is obtained by solving an eigenvalue problem. • The numerical results show good performance of the developed method. Fully Ceramic Micro-encapsulated (FCM) fuel is an important candidate for the accident tolerant fuel (ATF). Compared with traditional fuel, the double heterogeneity of FCM fuel makes the effective multi-group cross section calculation more challenging. In this paper, an improved disadvantage factor method is proposed to deal with the self-shielding effect of FCM fuel in the resonance energy range and non-resonance energy range, to achieve the equivalent homogenization of the FCM fuel. A new equivalent particle–matrix model was constructed by using the particle Dancoff factor to overcome the problem that the traditional volume-weight equivalent model could not consider the macro heterogeneity between fuel rods. Based on the new one-dimensional equivalent sphere model, the ultrafine group slowing down equation is solved to obtain the ultrafine group disadvantage factor in the resonance energy region. In the non-resonance energy region, the multi-group disadvantage factors of the fast group and thermal group are obtained by using the eigenvalue calculation in the new equivalent particle–matrix model. The proposed method has been implemented in the high fidelity neutronics program NECP-X and tested with a set of cases. The results show its good agreement with the Monte Carlo reference for both the reactivity and self-shielding cross sections. [ABSTRACT FROM AUTHOR]

[Display omitted] • Lateral and vertical 2H-1 T′ MoS 2 /ReS 2 heterojunctions were synthesized. • Growth mode is different for the high and low Re/S ratio. • Different interface structures in 2H-1 T′ MoS 2 /ReS 2 heterojunction were observed. • Heterojunctions exhibit weak PL spectra due to the type-I band structure. Two-dimensional (2D) 2H-1 T′ heterojunctions based on transition-metal dichalcogenides have attracted great attention due to their special interface structures and novel properties. Understanding the growth mechanism and the interface structures is essential for their controllable growth. In this work, we successfully synthesize lateral and vertical MoS 2 /ReS 2 heterojunctions by using a two-step chemical vapor deposition (CVD) method, and deeply explore the evolution of the structure. It is found that the growth mode of heterojunction undergoes a transition from lateral epitaxy to vertical stacking under a high Re/S ratio during the growth. Furthermore, the interface structure is greatly determined by the edge shape of MoS 2 core. Two kinds of structures, with the b-axis of 1 T′ ReS 2 shell parallel to the (1 0 0) and (0 1 0) plane of MoS 2 core respectively, are observed. Meanwhile, by comparing with MoS 2 monolayer, MoS 2 /ReS 2 heterostructures possess a reduced emission intensity due to the carrier transfer. This work provides a robust and controllable strategy for the synthesis of heterostructures with different phase structures. [ABSTRACT FROM AUTHOR]

• The development of neutron-photon-coupling transport in a newly developed Monte Carlo code NECP-MCX. • The neutron-photon-coupling analysis is performed for CFETR. • NECP-MCX is capable to obtain more reliable results of photon kerma than MCNP. The design of Chinese Fusion Engineering Test Reactor (CFETR) is being carried out. The neutronics analysis of the tritium-breeding blankets in CFETR is one of the essential tasks, and raises requirements for the neutronics-analysis codes such as the ability of simulating the neutron-photon-coupling transport and the ability of obtaining the accurate energy deposition of neutrons and photons. NECP-MCX is a hybrid Monte-Carlo-deterministic code newly developed by the Nuclear Engineering Computational Physics (NECP) Lab at Xi'an Jiaotong University. In order to meet the requirements raised by CFETR, the neutron-photon-coupling transport and other related functionalities are developed in NECP-MCX. The code-to-code verification shows that the results of NECP-MCX agree well with those of other codes. The application of NECP-MCX for the neutron-photon-coupling analysis of the tritium-breeding blankets in CFETR shows that NECP-MCX can obtain more reliable results of photon kerma than MCNP. [ABSTRACT FROM AUTHOR]

Tritium breeding capability is one of the most important objectives for fusion reactor design, and the tritium self-sufficiency has been set as a crucial target for the China Fusion Engineering Test Reactor (CFETR). The first wall (FW) is the first component of the blanket facing plasma, which bears high heat flux and high neutron current. Due to the neutrons entering into the blanket pass through the FW and then absorbed by the breeder materials, the FW design directly influences the blanket neutronics and tritium breeding performance. In this study, the helium-cooled solid breeder (HCSB) blanket of CFETR is selected as the reference design, and the impacts of the FW design parameters on tritium breeding performance have been studied comprehensively. A new conceptual mixed armor design and FW neutronics optimization guidelines have been proposed to enhance the tritium breeding ratio (TBR). Based on the mixed armor design and the optimization guidelines, the preliminarily optimized helium-cooled FW design shows about 9% TBR enhancement. • The concept of the mixed armor is proposed to provide better protection for the first wall and extra TBR enhancement. • The factors of the first wall design affecting the blanket neutronics performance are investigated in detail. • The first wall neutronics optimization guidelines for TBR enhancement are concluded. [ABSTRACT FROM AUTHOR]

To solve the deep-penetration problem in the Monte-Carlo simulation, the classic consistent adjoint driven importance sampling (CADIS) and forward-weighted CADIS (FW-CADIS) hybrid MC-deterministic method is adopted and enhanced in this paper. An automatic hybrid MC-deterministic code, NECP-MCX, is newly developed. NECP-MCX embeds the discrete-ordinate code NECP-Hydra to automatically perform the forward and adjoint neutron-transport calculations and set up the importance parameters for the MC simulation. In massive parallel calculation for large shielding problems, huge amounts of memory are required to save detailed weight-window information. To mitigate the memory limitation, a mesh-coarsening algorithm is developed based on the contributon theory. It can be applied in the traditional Cartesian-mesh and the nested-mesh structure. NECP-MCX was applied to the VENUS-3 benchmark. The results show that NECP-MCX with the enhanced hybrid method is effective in variance reduction and memory saving for the simulation of large and complicated shielding problems on parallel computer resources. [ABSTRACT FROM AUTHOR]

• The statistical data characteristics of dose rates at target points in the large shielding-design parameter space are analyzed. • Two commonly-used neural network-based methods are reviewed and their limitations in the large shielding-design parameter space are discussed through theoretical machine learning and data transformation analysis. • A new coupling approach S N -MscaleDNN approach is proposed to achieve rapid and accurate shielding-scheme evaluation in the large shielding-design parameter space. Rapid shielding-scheme evaluation in the large shielding-design parameter space (LSDPS) helps to optimize much better reactor shielding schemes. For this reason, a practical engineering approach utilizing a 1-D S N calculation and MscaleDNN network model is presented for accelerating reactor shielding-scheme evaluation in the LSDPS. We first review two commonly-used neural network-based methods for shielding-scheme evaluation and analyze their difficulties in the LSDPS through theoretical machine learning and data transformation analysis. Based on these discussions, the work transitions from method reviews to our two-step coupling solution to overcome these difficulties arising from high-frequency and multi-scale characteristics in the task. We first adopt equivalent 1-D S N calculations to quickly obtain inaccurate radiation dose rates at target points and then correct them by the frequency-scaled neural network MscaleDNN. The numerical results demonstrate that the proposed method achieves engineering-acceptable evaluation accuracy and higher accuracy than the conventional methods in the LSDPS. [ABSTRACT FROM AUTHOR]

• A non-iterative equivalence method is proposed to resolve the multi-group equivalence effect. • The new method is implemented into the resonance calculations of the Bamboo-Lattice code. • The new method guarantees less computational cost as compared to the SPH method. Multi-group heterogeneous reaction rates calculated by the subgroup method in resonance calculations are not reproduced, when subgroup-collapsed one-group homogeneous calculations are conducted for the same problem, giving rise to a multi-group equivalence effect. In this paper, a new non-iterative equivalence method introducing partial current discontinuity factors (PCDFs) is proposed to resolve this effect and employed into the Bamboo-Lattice code. Its validity was tested and verified through several fixed-source as well as eigenvalue problems for both single pin-cell and assembly geometries. The numerical results show that preservation of scalar flux, partial currents, neutron leakage and reaction rates is guaranteed by implementing this new method. Moreover, the computational time comparison for different equivalence methods shows that the newly proposed non-iterative equivalence method promises a significantly less computational cost compared to the traditional iterative super-homogenization (SPH) method in treating multi-group equivalence effect. [ABSTRACT FROM AUTHOR]

The China Fusion Engineering Test Reactor (CFETR) is a tokamak device to validate and demonstrate fusion engineering technology. In CFETR, the breeding blanket is a vital important component that is closely related to the performance and safety of the fusion reactor. Neutronics/thermal-hydraulics (N/TH) coupling effect is significant in the numerical analysis of the fission reactor. However, in the numerical analysis of the fusion reactor, the existing coupling code system mostly adopts the one-way coupling method. The ignorance of the two-way N/TH coupling effect would lead to inaccurate results. In this paper, the MCNP/FLUENT code system is developed based on the 3D-1D-2D hybrid coupling method. The one-way and two-way N/TH coupling calculations for two typical blanket concepts, the helium-cooled solid breeder (HCSB) blanket and the water-cooled ceramic breeder (WCCB) blanket, are carried out to study the two-way N/TH coupling effect in CFETR. The numerical results show that, compared with the results from the one-way N/TH coupling calculation, the tritium breeding ration (TBR) calculated with the two-way N/TH calculation decreases by −0.11% and increases by 4.45% for the HCSB and WCCB blankets, respectively. The maximum temperature increases by 1 °C and 29 °C for the HCSB and WCCB blankets, respectively. [ABSTRACT FROM AUTHOR]

• A high precision resonance calculation method is proposed. • The proposed method realizes the practical application of the hyperfine group method in the lattice and whole-core physics simulations. • The proposed method can obtain the correct spatially dependent effective cross sections. The hyperfine group method is considered as the most accurate resonance self-shielding method, but it suffers some limitations in the practical calculation due to its long computation time. In order to make it practical to use the hyperfine group method in large scale geometry problems, a cross section interpolation method is proposed in the frame of the newly developed global-local self-shielding method. In this paper, a series of typical 1-D pin cells are established by varying the Dancoff correction factor, burnup depth and fuel temperature. The hyperfine group method is just performed to these typical pin cells, and the cross sections of the realistic pin cells are obtained by interpolating according to the real values of these parameters. By this method, the times of the hyperfine group method needed to be performed is greatly reduced. This method has a very high accuracy in practical assembly and whole-core physics calculation. [ABSTRACT FROM AUTHOR]

The high-fidelity reactor physics calculation attracts a lot of attention because of the development of high performance computing technology and increasing requirement of reducing the approximation of traditional methods. Here introduces a new high-fidelity deterministic neutronics code NECP-X being developed at Nuclear Engineering Computational Physics (NECP) lab at Xi’an Jiaotong University. Compared to previous developed codes like DeCART, MPACT and nTRACER, it has its own separate features such as a pseudo-resonant-nuclide subgroup method, a new free-matrix CMFD acceleration method and an axial S N solver for improving accuracy, etc. The VERA progression problems 1 through 5-2D are tested to verify NECP-X, and the numerical results show that NECP-X can obtain accurate results for all of this verification problems. [ABSTRACT FROM AUTHOR]

The development and verification of a coupling code system named NECP-X/SUBSC which integrates a high-fidelity neutronics code NECP-X and a thermal-hydraulics (T/H) subchannel code SUBSC are presented. In order to accomplish high-fidelity, improved and realistic geometry modeling such as semi-explicit representation of grid spacer, resonance self-shielding treatment with pseudo-resonant-nuclide subgroup method, anisotropic scattering treatments and 2D/1D fusion method are implemented in the NECP-X code. A group of benchmark problems, including VERA core physics benchmark progression problems 2–3, are utilized to verify NECP-X. A sub-channel code SUBSC is developed for the pin-by-pin thermal-hydraulics calculation. The high-quality experimental data provided by the OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark is used to validate SUBSC. After assessing the two separate codes, an internal coupling method is used to integrate SUSBC with NECP-X. Finally, the coupling code system NECP-X/SUBSC is applied to the VERA core physics benchmark problem #6, 3D Hot Full Power (HFP) assembly. Axial normalized pin power shapes for the pin with the highest power and the pin at the assembly corner, subchannel exit coolant temperature and volume averaged fuel pin temperatures at the axial level of maximum temperature are compared to results from the Consortium for Advanced Simulation of Light Water Reactor (CASL)’s MPACT/CTF code. [ABSTRACT FROM AUTHOR]

Two effects in Fluoride-salt-cooled High-temperature Reactors (FHRs) are analyzed based on Monte Carlo simulation. Firstly, the thermal neutron scattering effect of fluoride salt (2LiF-BeF 2 ) is considered by utilization of the newly generated Thermal neutron Scattering Library (TSL) files. It is found that the neutron spectrum becomes harder and the fission reaction rate of 235 U decreases at thermal energy range due to the up-scattering introduced by thermal neutron scattering effect. Secondly, the resonance elastic scattering effect of heavy nuclides in epithermal energy range is covered by implementation of the Doppler Broadening Rejection Correction (DBRC) method in MCNP. It is shown that neutron up-scattering is enhanced in the low energy wing of the resonance peak and neutron down-scattering is increased in the high energy wing of the resonance peak. This phenomenon leads to an increase in the neutron capture rate of 238 U by about 1.0%. For the analyzed FHR pebble unit cells at 1200 K depending on TRISO packing factor, the thermal neutron scattering effect of 2LiF-BeF 2 and the resonance elastic scattering effect result in a decrease in k inf of 93–290 pcm and 131–591 pcm, respectively. By taking into account the two effects simultaneously, the k inf of FHR pebble unit cells decreases by 248–881 pcm. [ABSTRACT FROM AUTHOR]

The stochastic characteristics of the randomly distributed coated-fuel particles, the thermal-neutron scattering effect of the fluoride-salt and the resonance elastic scattering effect of heavy nuclides were neglected in early neutronics studies on the Pebble-Bed Fluoride-salt-cooled High-temperature Reactor (PB-FHR). In order to assess the impact of these effects on the neutronics calculation, the stochastic effect is analyzed by applying the explicit random modeling approach and Chord Length Sampling method, the thermal-neutron scattering effect of the fluoride-salt is quantified by evaluating and processing a new thermal-neutron scattering data library, and the resonance elastic scattering effect is covered by using the Doppler Broadening Rejection Correction (DBRC) method in this work. According to the critical calculations of the PB-FHR pebble unit cells with different TRISO Packing Factor (TPF), the different stochastic modeling methods lead to a difference of k inf by tens of pcm, the thermal scattering effect of 2LiF-BeF 2 results in a decreasement of k inf by 104–290 pcm and the resonance elastic scattering effect leads to a decreasement of k inf by 107–437 pcm, respectively. In addition, the thermal-neutron scattering effect and resonance elastic scattering effect can be addible, the total impact of the thermal scattering effect of 2LiF-BeF 2 and the resonance elastic scattering effect of heavy nuclide leads to a decreasement of k inf by 204–747 pcm. [ABSTRACT FROM AUTHOR]

A stepwise control strategy for enhancing glutathione (GSH) synthesis in yeast based on oxidative stress and energy metabolism was investigated. First, molasses and corn steep liquor were selected and fed as carbon source mixture at a flow rate of 1.5 g/L/h and 0.4 g/L/h, respectively, for increasing cell density in a 10 L fermenter. When the biomass reached 90 g/L, the KMnO4 sustained-release particles, composed of 1.5% KMnO4, 3% stearic acid, 2% polyethylene glycol and 3% agar powder, were prepared and added to the fermentation broth for maintaining the oxidative stress. The results showed that the maximum GSH accumulation of the group fed KMnO4 sustained-release particles was 39.0% higher than that of KMnO4-fed group. In addition to the improved average GSH productivity and average specific production rate, the activities of GSH peroxidase, γ-glutamylcysteine synthetase and GSH reductase, enzymes taking part in GSH metabolism, were also significantly enhanced by KMnO4 sustained-release particles feeding. Finally, 6 g/L sodium citrate fed as an energy adjuvant elevated the intracellular ATP level for further enhancing GSH production. Through the above stepwise strategy, the GSH accumulation reached 5.76 g/L, which was 2.84-fold higher than that of the control group. The stepwise control strategy based on oxidative stress and energy metabolism significantly improved GSH accumulation in yeast. [ABSTRACT FROM AUTHOR]

• The depletion capability of NECP-X is developed and validated in this work. The depletion capability of NECP-X is developed. Two depletion libraries are developed for depletion calculations. One full-fidelity depletion library contains 1547 isotopes, and a new compressed depletion library is generated. Cross sections from the EAF-2010 library are integrated into depletion libraries, which give cross sections for reaction channels not included in the ENDF library. For the traditional predictor-corrector method, two transport calculations are conducted during every depletion step, where transport calculation is very time consuming, so a new transport-depletion coupling strategy is developed to reduce calculation time. The VERA depletion benchmark problems are utilized to assess the accuracy of compressed depletion library and new coupling method. The equivalent pin-cell model and assembly model, which are often used for validation, are asserted by comparing to the full-core model. Finally, measurement data for 22 fuel samples are used to validate the depletion capability of NECP-X. The results demonstrate good accuracy of NECP-X. [ABSTRACT FROM AUTHOR]