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The research and development progress of NECP-MCX, a deep penetrating cross-scale radiation field analysis software independently developed by Xi'an Jiaotong University, is summarized. New methods and functions have been developed for new application scenarios such as large-space gamma ray radiation transport simulation, dose field simulation within containment, fusion reactor shutdown dose simulation and point source shielding problem. Precise calculation results of various problems and high performance indicate that NECP-MCX is able to cope with new challenges of the new application scenarios. The efficiency of consistent adjoint driven importance sampling (CADIS) and forward-CADIS (FW-CADIS) is shown in this work by calculation PCA-Replica benchmark which is a classic pressure vessel shielding problem in SINBAD (shielding integral benchmark archive and database). Both CADIS and FW-CADIS are able to improve simulation efficiency of reaction rate tallies in detectors. For the new challenge of large-space gamma ray radiation transport simulation, the CADIS-next event estimator (NEE) coupling method is implemented in NECP-MCX. By combining with CADIS, the convergence of NEE is improved as more paticles transport to the adjacent region of the detector point. With the coupling method, the point tally ability of CADIS is expanded as well, which allows CADIS method to obtain results at points. The CADIS-NEE method is able to accurately and efficiently obtain the photon flux density at km-scale distance in large-space gamma ray radiation transport simulation, and the computational efficiency is 6.8 times higher than that of the traditional NEE method. Further more, the relative standard deviation of CADIS-NEE decreases at a rate of ~N1/2 over the last half run, which means a reasonable convergence. While NEE's relative standard deviation is not able to decrease at such stable rate in the large-space gamma ray radiation problem. For the simulation of shut-down-dose-rate (SDDR) of China Fusion Engineering Test Reactor (CFETR), NECP-MCX can obtain and visualize decay photon strength and distribution of the whole CFETR after shutdown. Then SDDR of poloidal field coils, toroidal field coils, vacuum vessel and divertor at various shutdown time can be estimated by the novel particle-transport/depletion/activation/source-term coupling analysis process in NECP-MCX. The SDDR of CFETR from just shutdown to 500 years after shutdown is evaluated with NECP-MCX. For the point source shielding problem, NECP-MCX proposes the first-collision source (FCS)-CADIS method to cope with the limitation of source biasing of the CADIS method, and the computational efficiency of FCS-CADIS method is 2 times higher than that of the CADIS method.

The iterated fission probability (IFP) method and Contributon-linked eigenvalue sensitivity/uncertainty estimation via track-length importance characterization (CLUTCH) method are commonly used by many Monte Carlo codes for the sensitivity analysis of k-eigenvalue to continuous-energy nuclear data. The memory consumption of the IFP method is huge and the CLUTCH method is proposed to reduce the memory usage of the IFP method. However, it's hard to tally sufficiently-converged importance weighting functions for the mesh-based CLUTCH (CLUTCH-Mesh) method for large-scale problems, such as AP1000 whole core problem, subsequently reducing the accuracy of sensitivity coefficients calculated by the CLUTCH method. Therefore, the function expansion tally (FET) was used to tally importance weighting functions in the CLUTCH (CLUTCH-FET) method in this paper. Legendre polynomial was selected as the basic function of FETs and both segmented FETs with low orders and the global FET with the high order were implemented. During the tally process of FETs, the spatial locations of particles should be projected into domains of definition firstly, and then the tally scores of function expansion coefficients were obtained and accumulated. By averaging function expansion coefficients among inactive cycles, the FET was finished. The FETs were used for both ancestor fission neutrons distribution in original generations and the averaged progenies of the ancestor fission neutron in asymptotic generations. Based on the method mentioned above, the IFP, CLUTCH-Mesh and CLUTCH-FET methods were implemented in Monte Carlo code NECP-MCX. The verifications were conducted among Godiva, Flattop and AP1000 whole core problems by comparing the sensitivity results of CLUTCH-Mesh method, CLUTCH-FET method and the IFP method, whose results were selected as the reference. The numerical results indicate that for small-scale problems, such as Godiva and Flattop problems, the CLUTCH-Mesh and CLUTCH-FET methods are able to calculate the sensitivity results with the same high precision as the IFP method. In addition, the tally efficiency of CLUTCH-Mesh method equals to that of the CLUTCH-FET method and the tally efficiency of these two methods are significantly higher than that of the IFP method. For large-scale problems, such as the AP1000 whole core problem, large bias exists between the sensitivity results of the CLUTCH-Mesh method and those of the IFP method because very large number of particles are needed to obtain sufficiently-converged importance weighting functions. However, the CLUTCH-FET method remains high accuracy and tally efficiency for large-scale problems. The FOM (figure of merit) values of the CLUTCH-FET method can be increased by 5.2 and 6.0 times when compared to those of the IFP method and the CLUTCH-Mesh method, respectively. Therefore, the CLUTCH-FET method is recommended for sensitivity calculation of large-scale problems. Segmented FETs with low orders are more recommended than a global FET with high order to avoid under-fitting or over-fitting problems.

There are various geometric fuels in reactors, including cylindrical fuel such as SPERT test reactor, plate fuel such as JRR3 research reactor, annular fuel such as Xi’an Pulse Reactor, and other geometric fuel such as advanced test reactor, etc. Many researchers made efforts for the resonance calculation of these geometric kinds of fuels and several methods were proposed, such as the equivalence theory, the ultrafine group method, the subgroup method, and the globallocal method. At the beginning of the equivalence theory and the ultrafine group method, only cylindrical fuels and plate fuels are supported. Thus, some improvements were proposed in the following years, such as trinomial rational approximation for the equivalence theory and the MOC (method of characteristics) based ultrafine group method. However, these methods have some accuracy and efficiency problems. As for the subgroup method, although arbitrary geometry fuel can be treated since the subgroup fixed source equation can be solved by the MOC, the geometry processing capability of the subgroup method is limited by the resonance integral table. As for the uniformity problembased resonance integral table, the accuracy is not enough. While for the nonuniformity problembased resonance integral table, the arbitrary geometric ability will be lost. For the globallocal method, only cylindrical fuel and plate fuel can be solved. Thus, in this paper, based on the globallocal coupling strategy, a resonance calculation method based on the equivalent geometry method for complex geometric fuels was proposed. First, for the isolated system of complex geometry fuel, based on the conservation of escape probability, an equivalent onedimensional (1D) cylindrical (or plate) fuel model of the complex geometry fuel was established. Secondly, based on the conservation of the collision probability of the fuel to the outer structure material, the equivalent size of the fuel outer structure material was obtained. Then, based on the Dancoff factor conservation of the complex geometry fuel, the equivalent moderator of 1D cylindrical (or plate) fuel model was established. And finally, for the equivalent 1D cylindrical (or plate) model, the pseudoresonantnuclide subgroup method was used to calculate the selfshielding crosssection. The method was applied to the resonance calculation of noncylindrical geometry fuel. As for the pin cases and assembly cases, the deviation of the microscopic absorption section is less than 320%, and the deviation of the eigenvalue is less than 150 pcm. The power deviation is less than 0128%. Compared with the MOCbased pseudoresonantnuclide subgroup resonance calculation method, under the comparable calculation accuracy, the calculation efficiency increases by more than two orders of magnitude. For the core level case, the eigenvalue difference is 338 pcm, the maximum deviation of assembly power is 1182%. The results show geometry flexibility, high accuracy, and high efficiency of this method.

AP1000 is the typical third-generation nuclear power technology. The sensitivity analysis of continuous-energy nuclear data for AP1000 is the basis of uncertainty quantification which plays an important role in subsequent safety analysis for AP1000. Multi-group approximations exist in sensitivity analysis based on multi-group energy nuclear data. Therefore, it’s necessary to calculate sensitivity coefficients based on continuous-energy nuclear data. In this paper, adjoint flux was calculated based on iterated fission probability (IFP) in Monte Carlo forward calculation and then sensitivity coefficients of keff to nuclear data were obtained through first-order perturbation theory. The sparse-matrix method was used to reduce the large memory usage in IFP, while the overlapping block method was used to improve counting efficiency and large statistics fluctuation. The function module of sensitivity analysis was developed in Monte Carlo code NECP-MCX. The module developed above was verified based on the VERA 2B problem. Both energy-dependent and energy-independent sensitivity coefficients calculated by NECP-MCX are all within 3σ of those obtained by MCNP, which indicates the accuracy of the NECP-MCX. In addition, the sensitivity coefficients related to the scattering cross section fluctuate greatly, which is because when the magnitudes of the scattering term and the total term in the sensitivity coefficient formula are the same, the subtraction of these two terms will result in a value with a larger statistical fluctuation. Furthermore, the sensitivity analysis of continuous-energy nuclear data was performed with enhanced NECP-MCX for AP1000 and the nuclear data that are most sensitive to keff were analyzed, which are 235U-ν, 235U-σf, 235U-σt, 238U-σγ, 1H-σelas and 10B-σt, where 235U-ν, 235U-σf and 235U-σt are mainly related to the production of neutron, 238U-σγ and 10B-σt are dominant in the disappearance of neutron and 1H-σelas is dominant in the slowing down of neutron. In this paper, the sensitivity coefficient of keff to nuclear data was derived based on first-order perturbation theory. The module of continuous-energy nuclear data sensitivity analysis was implemented in Monte Carlo code NECP-MCX. The verification of the module was conducted based on VERA 2B problem. Finally, the sensitivity coefficients of keff to continuous-energy nuclear data were calculated for AP1000. The most sensitive nuclear data to keff were obtained from the results. In the future, based on the obtained sensitivity coefficients and the covariance information in the nuclear data, the uncertainty of the AP1000 core keff calculation caused by nuclear data can be obtained through sandwich rule for subsequent safety analysis.

Chinese Fusion Engineering Test Reactor (CFETR) has been proposed to bridge the technology gap between the International Thermonuclear Experimental Reactor (ITER) and the Fusion Demonstration Reactor (DEMO). As the most crucial nuclear component, fusion blanket undertakes the functionalities of tritium breeding, energy extraction and radiation shielding. The design of fusion blanket directly determines whether the fusion reactor can operate in a safety and steady way. In the design of fusion blanket, the neutronics and thermal-hydraulics are the most associated aspects. Actually, the neutronics and thermal-hydraulics are closely coupled to each other in the fission reactor. The effect caused by the interaction between neutronics and thermal-hydraulics is called the neutronics/thermal-hydraulics coupling effect. In the fission reactor, the neutronics/thermal-hydraulics coupling effect will lead to significant impact on both of neutronics and thermal-hydraulics, e.g., change the effective multiplication factor, and influence power distribution and temperature distribution. Therefore, the neutronics/thermal-hydraulics coupling effect must be taken into account in the design work of fission reactor. However, the influences of the neutronics/thermal-hydraulics coupling effect are still not clear in the fusion reactor. The key point to perform the neutronics/thermal-hydraulics coupling calculation of fusion blanket is to choose rational solvers and coupling method. Considering the complicated structure of the fusion blanket, Monte-Carlo code MCNP and computational fluid dynamics code FLUENT, which have good geometric adaptability, are selected as the neutronics solver and the thermal-hydraulics solver, respectively. Nevertheless, the direct three-dimensional coupling calculation is still difficult because of the complicated geometric mapping relationship and huge amount of calculation. Thus, a hybrid 3D-1D-2D coupling method is used to deal with the spatial mapping between neutronics model and thermal-hydraulics model. Moreover, the pseudo material method is adopted to efficiently handle the variation of neutron cross sections under different temperatures. In this paper, the main conceptual blanket designs of CFETR, including the helium cooled solid blanket design under 200 MW, and the water cooled solid blanket designs under 200 MW and 1.5 GW, were selected as the research objects to study the neutronics/thermal-hydraulics coupling effect on the CFETR. Above all, the neutronics sensitivity analysis of material temperature of fusion blanket was conducted to demonstrate how temperature influenced the tritium breeding capability and nuclear heat deposition. The results of sensitivity analysis indicate that the thermal scattering effect of beryllium and the density of water are the decisive factors of temperature effect. Subsequently, the neutronics/thermal-hydraulics coupling calculation was carried out. The coupling results show that, in the helium cooled solid blanket, the neutronics/thermal-hydraulics coupling effect is small and can be ignored, and in the water cooled solid blanket, the neutronics/thermal-hydraulics coupling effect will slightly influence the tritium breeding capability and temperature distribution.

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]