[en] Reactor pressure vessel is the key equipment in making correct decisions on design plant lifetime. Changes in mechanical properties of reactor vessel materials result from exposure to fast neutron. The use of MOX fuel in LWRs present different neutron characteristics, whether the software we used can calculate the structural integrity of reactor components with high degree worth research. This paper use MCNP, TORT, SCALE to calculate VENUS-2 benchmark the calculation shows that all this software can get reasonable result that can be used in design. MCNP has the highest accuracy. (authors)

[en] Based on the adjoint transport technology, the ex-core detector response function under the full power operation of one third-generation reactor is calculated by using the two-dimension SN code. Calculation results are compared with the original data and the radial weighting factors would not exceed 3%, and the axial weighting factors would not exceed 1%. The result of this calculation proves that the calculation model and method is correct. The technology has been validated in the Ling'ao Phase II Nuclear Power Plant Compared to the numerical result of the radial weighting factors via SN forward Transport Method, it shows that the calculation results difference of thc two methods would not exceed 5%. It proves that the result for the two methods is identical and the computational efficiency of the Adjoint Transport Method is more efficient. (authors)

[en] Based on the genetic algorithm, this paper focuses on the optimization of the nuclear reactor shielding design and conducts the studies about the single-objective and multi-objectives optimization problem of the reactor shielding design. The methods, developed in this paper, have been validated by the shielding of the Savannah nuclear ship. The effectiveness and correctness of the single-objective and multi-objectives optimization methods for nuclear reactor shielding have been fully demonstrated. A new technique has been provided for the shielding optimization of the reactor design in the future. (authors)

[en] Based on rigorous two-step (R2S) method, a three-dimensional shutdown dose rate calculation code was developed for nuclear devices with large dimensional and complex geometries, which integrated the functions of neutron transport calculation, activation calculation and decay gamma transport calculation by automatic coupling the Monte Carlo particle transport calculation code MCNP with the activation simulation code FISPACT. The code was applied to the shutdown dose rates analysis of EAST (Experiment Advanced Superconducting Tokamak), and the three-dimensional dose map of EAST can be used to assist radiation protection of EAST. (authors)

[en] A coupling calculation method based on the discrete ordinate method code DOT and Monte-Carlo method code MCNP was realized. By automatically providing a binary source specification file for the surface source function of MCNP (i.e. SSR command), this coupling method was achieved and subsequently the necessity of re-compiling of MCNP code was eliminated. This coupling method was applied in the shielding problem of rooms around the pit of reactor cavity and a good agreement was obtained between the calculation and measurement results, which validated the correctness of this coupling method. Further more, source bias method was implemented in this coupling method, which shows great efficiency in variance reduction and the improvements of calculation efficiency. (authors)

[en] The post-processing capability of the current MCNP is not enough to effectively and sufficiently analyze calculation results of complex problems. In the paper, a geometry coupled visual analysis system for MCNP was self-developed in FDS Team by using scientific visualization technology, GPU (graphics processing unit)visual programming technology, and the reversion engine of MCAM. The visual analysis system was implemented to support data extraction and manipulation based on graphics user interface, three-dimensional plotting of tally results, and visual analysis of the results with geometries. And the system has been tested by many de facto cases, such as ITER. Testing results showed that the system provided an intuitively and effectively visual analysis approach for the post-processing of MCNP and obviously enhanced the analysis efficiency. (authors)

[en] MCNP is a general and powerful Monte Carlo code for photons, neutrons, and electrons transport simulation and is widely used in nuclear analysis. However, external specialized data analysis codes may be needed to analyze complex MCNP calculation results. Especially, there is a lack of efficient data analysis codes which supports geometry-coupled visual analysis. To address the post-processing concerns, a geometry-coupled visual analysis system for MCNP, was developed with 3D scientific visualization and graphics processing unit programming technologies by FDS Team. It was designed and implemented to provide friendly graphical user interface based data manipulation, 3D plotting of tally results, and visual data analysis coupled with geometries. Testing and preliminary application results showed that the system provided a satisfied visual analysis approach for the post-processing of MCNP and obliviously enhanced the analysis efficiency. (author)

[en] Highlights: • Systemic activation analysis of FLiBe has been firstly carried out for the molten salt reactor. • The activation analysis was carried out based on an actual reactor design, TMSR-SF1. • The nuclides for shielding design, occupational exposure and environment have been studied. • The initial impurities and corrosion products during operation have both been considered. - Abstract: As an excellent material for candidate coolant of the molten salt reactor, the activation of 2LiF-BeF₂ is an important issue and should be concerned. In this paper, the activation of 2LiF-BeF₂ has been analyzed based on a specific Molten Salt Reactor. And combined approaches, Monte Carlo method for neutron transport and extension of the Euler for inventory calculation, have been adopted. The short life nuclides for shielding design, the long-life nuclides for occupational radiation exposure management and ³H, ¹⁴C production for environment protection have been studied.

[en] MCNP is a general and powerful Monte Carlo code for photons, neutrons, and electrons transport simulation and is widely used in nuclear analysis. However, external specialized data analysis codes may be needed to analyze complex MCNP calculation results. Especially, there is a lack of efficient data analysis codes which supports geometry-coupled visual analysis. To address the post-processing concerns, a geometry-coupled visual analysis system for MCNP, was developed with 3D scientific visualization and graphics processing unit programming technologies by FDS Team. It was designed and implemented to provide friendly graphical user interface based data manipulation, 3D plotting of tally results, and visual data analysis coupled with geometries. Testing and preliminary application results showed that the system provided a satisfied visual analysis approach for the post-processing of MCNP and obviously enhanced the analysis efficiency. (author)

[en] The coupling calculation of transport and depletion usually adopts traditional predict-correct (PC) burnup method in PWR fuel assembly. However, the method results in the calculation error because of its assumption. In order to improve the accuracy of burnup calculation, the modified PC burnup method was studied and high-order PC burnup method which corrected the microscopic reactivity was improved. Corresponding code was developed to verify and analyze these burnup methods with some benchmarks. The results indicate that new methods improve the calculation accuracy under the premise of insuring calculation efficiency. (authors)