[en] Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

[en] In this report we present the results of an investigation in developing the core loading pattern of the VVER-1200/V491. The loading pattern was developed in two ways: (1) Based on the core configuration of the VVER-1000/V392 and (2) Selected by the optimized method using in the LPO-V code. To estimate the developed core, several parameters such as the k_eff, power distribution, delayed neutron fraction, fuel and moderator temperature reactivity feedbacks are calculated by SRAC code. (author)

[en] In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

[en] The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant. However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER-1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code.To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel and moderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR. (author)

[en] The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods. (author)

[en] This investigation aims at considering the effect of the coolant density change on the power distribution of a VVER-1000 fuel assembly based on building a coupling calculation scheme using two codes MCNP5 and COBRA-EN. The origin of the coolant density change is due to the neutronic-thermal hydraulic feedback. The master program for the coupling calculations was written using PERL, where the input is built for MCNP5 and COBRA-EN. In this work the results of two successive iterations were compared and it shows that the convergence was obtained after the 3rd iteration with 50 millions of histories to be performed in the MCNP calculations when the convergence criterion is set at 1 percent. (author)

[en] Small Modular Reactors (SMR) and their applications are receiving more attention from the nuclear community. One of the applications is the use in floating nuclear power stations, however, floating nuclear power stations are potentially at risk and requires higher level of safety. In addition to the combination of active and passive safety systems, the use of accident tolerant fuel (ATF) by improving or replacing fuel ceramic materials and use other fuel cladding materials has been studied in order to improve the safety. As a result, the operators will has more time to respond to events, that could result in the release of fuel material and fission products. The project focuses on studying and designing a fuel with high burn-up for the ACPR50S reactor - the type of reactor expected to be installed on China's floating nuclear power stations.

[en] Neutronic characteristics for a VVER-1000 were investigated by using SRAC code and nuclear data library JENDL-3.3 with 107 public energy groups. The elementary lattice modules, PIJ and CITATION, have been used for modeling of the fuel rods, fuel assemblies and full core. The main Neutronic characteristics analyzed in this work include infinite multiplication factors (k_inf) versus burnup, the distribution of nuclide concentrations in the pin cells; the pin-wise power distribution in the assembly; the effective multiplication factors (k_eff), and the power distribution in the core. (author)

[en] Neutronic characteristics of a Japanese PWR 900 MWe were investigated by using SRAC code and nuclear data library JENDL-3.3 with 107 public energy groups. The lattice modules, PIJ and CITATION, have been used for modeling the fuel rods, fuel assemblies and full core. The main neutronic characteristics analyzed in this work include infinite multiplication factors (k_inf) versus burnup, the distribution of nuclide concentrations in the pin cells; the pin-wise power distribution in the assemblies; the effective multiplication factor (k_eff), and the power distribution in the core. (author)

[en] Machine learning (ML) using artificial neural network (ANN) methods is being strongly applied in nuclear reactor research. It is used to predict required parameters based on learning from big data sets. The ML models usually give much faster calculation speed while the accuracy is good and can be even better than the normal calculation tools. In this work, a Multi-layer perceptron network was built and trained to predict k-eff and peaking factor of a small modular PWR core. The obtained results are compared with the SRAC2006 code to evaluate the performance of the models. The comparisons show good accuracy and performance of the models. (author)