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[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] 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] 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] 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] 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 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] With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive especially in criticality analysis of initial core and shielding calculations, due to its advantages of flexible geometry modeling and use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with different physics and feedback interactions and coupling. To perform the high fidelity multi-physics simulations of real reactors or benchmark calculations such as two-cycle BEAVRS benchmark based on measurement data of a practical nuclear power plant, several factors must be considered such as large scale detailed depletion, thermal-hydraulics feedback, on-the-fly nuclear cross section processing, criticality search and inter-cycle refueling. In this paper, the abilities mentioned above for multiple burnup cycles simulations in Hot Full Power condition of PWR full core have been developed in continuous-energy Reactor Monte Carlo neutron and photon transport code RMC, which is developed by Department of Engineering Physics at Tsinghua University, Beijing (Wang et al., 2015). RMC has the capacity for lifecycle simulations of nuclear reactor cores. The BEAVRS benchmark was selected as an example and RMC was applied to a full core, two cycle burnup calculation of BEAVRS. All of the parameters given in the BEAVRS benchmark have been calculated and compared with the measured values of BEAVRS benchmark. For other parameters such as pin power distributions, they are compared to the results of other codes. The results of RMC agree well with the measured values of BEAVRS benchmark and also agree well with those of other codes. This was the first time for a Monte Carlo code to perform the full core, two cycle calculation of BEAVRS. This work paves the way for Monte Carlo codes in life cycle simulations of nuclear reactor cores.
[en] Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.
[en] The 2-D/1D whole-core transport method has been widely studied for 3-D fine flux or power distributions and implemented in many deterministic codes. KYCORE, a 2-D/1-D transport code, developed a new iteration strategy to calculate forward flux. In order to perform kinetics parameter computation and sensitivity analysis, an adjoint flux solution is required in modern reactor physics analysis. In this study, a new adjoint neutron transport solver, named KYADJ, was developed by utilizing the 2-D MOC and 1-D SN coupling method. A lattice test problem and C5G7 OECD/NEA 3-D benchmarks were used to verify the forward-adjoint neutron transport calculation of KYADJ. Forward-adjoint multiplication factors, forward flux, adjoint flux and kinetics parameters were compared with reference results generated by the Reactor Monte Carlo code RMC. Results showed that KYADJ agreed well with RMC and KYADJ had the ability to provide fine pin-by-pin forward and adjoint flux distributions and accurately compute kinetics parameters.
[en] Highlights: • Sensitivity analysis based on CPT is applied to the 2-D/1-D whole-core transport code KYADJ. • 2-D MOC and 1-D Sn coupled forward-adjoint neutron transport equations are solved by an advanced iteration approach. • A new covariance data file is processed from ENDF-B/VII.1 for uncertainty quantification. - Abstract: An evaluated nuclear data file is one of the important footstones in reactor physics design. Nuclear data can be viewed as arguments of output parameters. Therefore, the uncertainty of nuclear data can directly influence on the uncertainty of output parameters. A process of Sensitivity and Uncertainty (S&U) analysis is used to quantify the influence. In this study, we choose 2-D/1-D transport code to develop S&U analysis in order to reduce the impact of numerical model uncertainty. We can solve 2-D/1-D adjoint transport equation and apply the classical perturbation theory to sensitivity analysis of eigenvalue with respect to nuclear data. In addition, we obtain the uncertainty of eigenvalue using the “Sandwich” rule. The Peach Bottom-2 (PB-2) BWR cell benchmark and the Three Mile Island-1 (TMI-1) PWR cell benchmark are applied to verification. The Reactor Monte Carlo code RMC is used as a reference code. The results show that the code has ability of eigenvalue sensitivity and uncertainty analysis with high accuracy.