[en] In reactor analog simulation, the Monte Carlo method has relatively high fidelity since it can reduce calculation errors resulting from using approximation methods by describing complex geometry and neutron energy spectra. The Monte Carlo method is, in essence, a stochastic simulation of the neutron transport process, and statistical uncertainty of all the computed results must be provided. In large-scale reactor computation, this method can accurately estimate integral quantities such as the effective multiplication factor (K_eff), and it can limit the uncertainty within an acceptable range via reasonably selecting simulation particles. Nevertheless, it faces great challenges in estimating local quantities like neutron flux and fission power, with uncertainty of estimators beyond acceptance. To decrease the uncertainty of local quantities by increasing simulation particles may cost much time and reduce efficiency. The main reasons are as follows. Firstly, the scale of reactors is so large that common pressurized water reactors have tens of thousands of lattice cells; secondly, uneven distribution of neutron flux and fission power causes uneven distribution of uncertainty. Smith K. puts forward the 95/95 principle to estimate whether uncertainty calculated with Monte Carlo Code satisfies the requirement. According to the principle, with a confidence level of 95%, statistical errors of power density in more than 95% of the regions should be less than 1%. To meet the requirement of the principle, the uniform fission site (UFS) method is raised currently. UFS aims at redistributing fission sources to uniformize the fission source sampling, which can make statistical variance distribution more uniform. This method, which is easy to use, will not obviously increase computation time. Many studies show that UFS has a good variance reduction effect in Monte Carlo criticality calculation, but less attention is paid to its performance in burnup calculation and thermal-hydraulic coupled calculation. Uncertainty arising from the Monte Carlo method has a large impact on burnup calculation and thermal-hydraulic coupled calculation. Variance propagates between burnup steps and between iterative steps of coupled calculation, which causes asymmetry of results of the physical symmetrical regions and instability of coupled iterative convergence errors. To further verify the effect of UFS, this study uses it in the RMC (the Reactor Monte Carlo code) burnup calculation module and the neutronics/thermohydraulics coupled calculation module, and conducts two-dimensional pressurized water reactor burnup calculation and BEAVRS benchmark thermalhydraulic coupled calculation. In this way, it verifies the improvement effect of UFS on variance propagation. (authors)

[en] Objective: To observe the therapeutic effect and short-term reaction of radiotherapy synchronized with oral Hong Dou Shan chemotherapy for treatment of esophageal carcinoma. Methods: From 1997 to 2001, 50 patients with esophageal carcinoma were treated with radiotherapy synchronized with Hong Dou Shan capsule chemotherapy (group A) and another 50 such patients treated with radiotherapy alone (group B). Clinical therapeutic effects was compared between these two groups. Results: The patients were followed-up for three years. The 1-, 2-, and 3-year survival rates were 76%(38/50), 68%(34/50) and 58%(29/50) in group A and 56% (28/50), 44%(22/50) and 38% (19/50) in group B, respectively (P<0.05). Improvement of the clinical symptoms was more evident in group A than in group B. Conclusion: Radiotherapy synchronized with Hong Dou Shan capsule chemotherapy for treatment of esophageal carcinoma could increase the tumor extinctive rate and short-term survival rate without increase of adverse reactions. Hong Dou Shan is one of the ideal chemical drugs when synchronized with radiotherapy for treatment of esophageal carcinoma. (authors)

[en] To perform the generalized sensitivity analysis for nuclear data in a reactor physics design code, KYLIN-II, the generalized perturbation theory is adopted and several generalized fix-source equations with the orthogonal definite condition need to be solved when the sensitivity coefficients are figured out. Besides, the paper develops a new approach, CMFD-based generalized fix-source equation solution, to accelerate the convergence. The convergence efficiency of the generalized fixed-source equation is improved by roughly 4.3 times, and the sensitivity coefficients calculated by the GPT accord with those calculated by the direct perturbation theory, which demonstrates the sensitivity analysis ability in KYLIN-II. (authors)

[en] Evaluated nuclear data libraries are the basis of reactor physics analysis. This study investigates major versions of evaluated nuclear data libraries in various stages, namely ENDF/B-Ⅳ, Ⅴ.2, Ⅵ.8, Ⅶ.0 and Ⅶ.1, which are then processed by the internationally renowned nuclear data processing code NJOY to obtain five groups of continuous energy point cross section libraries. Micro cross sections of certain nuclides are compared and Reactor Monte Carlo Code (RMC) is used for verification of criticality benchmark. The result shows that continuous energy neutron cross section libraries, based on ENDF/B-Ⅶ.1, has higher accuracy and reliability. (authors)

[en] The development of advanced thorium-based nuclear system raises new requirements on nuclear data. The multi-group data file of critical nuclides in the thorium-uranium recycle is the foundation of physical design, analysis and calculation of the reactor core. Based on authoritative nuclear data processing code NJOY, this paper obtains a WIMS format multi-group cross section data files through processing the ENDF/B-VII.1 evaluation nuclear data file, uses the specific update maintenance procedure WILLIE to get a WIMS format data file, and conducts a series of critical benchmarks on the data file using the multi-group reactor core calculation code WIMSD5B. The results show that the computed results of the WIMS file based on the processing of ENDF/B-VII.1 are basically the same as those of the latest WIMS-D file published on the websites of the 'WIMS-D' library updating project (WLUP) with higher accuracy and reliability than those of the shipped WIMS-D file of the WIMSD5B code. Furthermore, the average deviation of the new WIMS file performing in the validation of 16 thorium-uranium cycle benchmarks is 0.225 3% smaller than that of the old WIMS file. (authors)

[en] Reactor Monte Carlo code (RMC) was constantly developed for reactor large-scale precise analog calculation. In criticality calculation, the maldistribution of reactor neutron flux and power density lead to greater volatility in statistical bias, causing asymmetries in the computed results of reactor physical symmetry zones. Computation asymmetries in Monte Carlo codes are mainly caused by statistical variance, which can be reduced through decreasing variance volatility. In this study, uniform-fission-site method was added into RMC criticality calculation, with Hoogenboom-Martin benchmark verified by calculations. The result shows that the modified algorithm can significantly reduce the variance volatility of reactor cores and the variance of low power regions. (authors)

[en] Adjoint sensitivity analysis (ASA) and the forward sensitivity analysis (FSA) based on the reduced order module (ROM) were developed on the 2D/1D transport code KYCORE. Sensitivity analysis on the TMI-1 PWR cell benchmark was performed. Results show that sensitivity coefficients of the effective multiplication factor (K_eff) with respect to ²³⁵U fission and capture cross sections calculated by KYCORE and RMC agree well with each other. However, the dependency exists in the resonance region. Sensitivity coefficients calculated by the FSA based on the ROM and the direct numerical perturbation (DNP) are consistent. Therefore, ASA and the FSA based on the ROM developed on the 2D/1D transport code KYCORE are verified. (authors)

[en] Based upon advances in theoretical algorithms, modeling and simulations, and computer technologies, the rational design of materials, cells, devices, and packs in the field of lithium-ion batteries is being realized incrementally and will at some point trigger a paradigm revolution by combining calculations and experiments linked by a big shared database, enabling accelerated development of the whole industrial chain. Theory and multi-scale modeling and simulation, as supplements to experimental efforts, can help greatly to close some of the current experimental and technological gaps, as well as predict path-independent properties and help to fundamentally understand path-independent performance in multiple spatial and temporal scales. (topical review)

[en] Highlights: • The physics adjoint and the mathematical adjoint are compared based on the 2-D/1-D solver. • Different adjoint flux functions are used to perform S&U analysis for comparison. • The SF96 problem and the PB-2 assembly problem are selected for verification. - Abstract: Sensitivity and Uncertainty analysis (S&U) based on the Classical Perturbation Theory (CPT) requires the solution of the adjoint flux. The adjoint flux is used as a weight function for sensitivity calculation. In the previous study, the MOC-based adjoint flux in the 2-D/1-D transport solver was given and applied to S&U analysis in the UAM benchmarks. The eigenvalue $k$ obtained by the forward calculation is different from that obtained by the adjoint calculation. In order to illustrate the phenomenon, three different adjoint flux functions, the MOC physics adjoint, the CMFD physics adjoint, and the CMFD mathematical adjoint are solved based on the 2-D/1-D solver. Next, three different adjoint flux functions are used to perform S&U analysis. The SF96 problem and the PB-2 assembly problem are selected for verification. Recommendation is made for the proper application of the CMFD mathematical adjoint flux to sensitivity calculation due to less computation time.

[en] Nuclear data sensitivity analysis and uncertainty propagation have been extensively applied to nuclear data adjustment and uncertainty quantification in the field of nuclear engineering. Sensitivity and Uncertainty (S&U) analysis is developed in the KYADJ whole-core transport code in order to meet the requirement of advanced reactor design. KYADJ aims to use two-dimension Method of Characteristic (MOC) and one-dimension discrete ordinate (S_N) coupled method to solve the neutron transport equation and achieve one-step direct transport calculation of the reactor core. Developing sensitivity and uncertainty analysis module in KYADJ can minimize deviations caused by modeling approximation and enhance calculation efficiency. This work describes the application of the classic perturbation theory to the KYADJ transport solver. In order to obtain uncertainty, a technique is proposed for processing a covariance data file in 45-group energy grid instead of 44-group SCALE 6.1 covariance data which is extensively used in various codes. Numerical results for Uncertainty Analysis in Modelling (UAM) benchmarks and the SF96 benchmark are presented. The results agree well with the reference and the capability of S&U analysis in KYADJ is verified.