[en] Highlights: • A new dynamics code for MSR named NTH3D-MSR has been developed. • The Predictor-Corrector Improved Quasi-Static and Nodal Green’s Function Method were implemented in NTH3D-MSR. • The NTH3D-MSR has the ability to carry out coupled neutronics and thermal-hydraulics calculation. - Abstract: The special characteristic of the Molten Salt Reactors (MSR) is that the liquid molten salt mixtures, acting like a coolant and fuel, are flowing through the core and the primary circuit. The delayed neutron precursors are driven by the fuel flow, which results in unique neutron dynamics characteristics of the MSR. The coupled neutronics and thermal-hydraulics code for MSR named NTH3D-MSR has been developed in this paper to simulate the dynamics characteristics of the MSR. In this code, the Predictor-Corrector Improved Quasi-Static (PC-IQS) method and the Nodal Green’s Function Method (NGFM) were adopted to solve the neutron kinetics equation and neutron diffusion equation respectively, and the multi-channel thermal hydraulic model is developed to consider the thermal feedback. In addition, the few-group constants of the graphite-moderated channel-type molten salt reactor are generated by the continuous-energy Monte-Carlo code Serpent which can provide remarkable geometry flexibility against the deterministic lattice transport codes. In order to verify the NTH3D-MSR code, the experimental data collected in the European project MOST are adopted to validate the coupled neutronics and thermal-hydraulics model of the NTH3D-MSR code. The numerical results indicate that the Serpent/NTH3D-MSR code has the capability for the dynamics analysis for the graphite-moderated channel-type molten salt reactors.

[en] Highlights: • New data libraries are generated to quantify the thermal neutron scattering effect. • DBRC is implemented in MCNP to quantify the resonance elastic scattering effect. • The impacts of the two negative effects does not interfere each other. • A significant improvement of critical calculation has been made for the FHRs. - Abstract: Two effects in Fluoride-salt-cooled High-temperature Reactors (FHRs) are analyzed based on Monte Carlo simulation. Firstly, the thermal neutron scattering effect of fluoride salt (2LiF-BeF_2) is considered by utilization of the newly generated Thermal neutron Scattering Library (TSL) files. It is found that the neutron spectrum becomes harder and the fission reaction rate of "2"3"5U decreases at thermal energy range due to the up-scattering introduced by thermal neutron scattering effect. Secondly, the resonance elastic scattering effect of heavy nuclides in epithermal energy range is covered by implementation of the Doppler Broadening Rejection Correction (DBRC) method in MCNP. It is shown that neutron up-scattering is enhanced in the low energy wing of the resonance peak and neutron down-scattering is increased in the high energy wing of the resonance peak. This phenomenon leads to an increase in the neutron capture rate of "2"3"8U by about 1.0%. For the analyzed FHR pebble unit cells at 1200 K depending on TRISO packing factor, the thermal neutron scattering effect of 2LiF-BeF_2 and the resonance elastic scattering effect result in a decrease in k_i_n_f of 93–290 pcm and 131–591 pcm, respectively. By taking into account the two effects simultaneously, the k_i_n_f of FHR pebble unit cells decreases by 248–881 pcm.

[en] Transient characteristics and flows in a centrifugal pump during its starting period were experimentally and numerically investigated. The two-dimensional particle image velocimetry technique was used to capture transient flow evolutions in the pump's diffuser. A new dynamic slip region method that combines the dynamic mesh method with the non-conformal grid boundaries is proposed to resolve the transient flows caused by the started impeller. Numerical self-coupling was realized by establishing a circulation pipe system along with the pump model equivalent to the experimental pump system. Numerical and experimental results agree well in both explicit characteristics and internal transient flow structures, confirming the validity of the proposed method. Analysis of the instantaneous flow in the impellers indicates that for the early stage of the startup, the transient vortex evolution between blades is the main reason for the transient head coefficient being lower than the steady state value. The reversed flow at the blade inlet is a less important reason for this effect. In later stages, the weakening of the intensity of the spatial vortex visible on S_2m and the main flow stream are the main reasons for the explicit performance slowly rebounding to the steady value

[en] An experimental study has been carried out in order to analyze the cavitation of a centrifugal pump and its effect on transient hydrodynamic performance during transient operation. The transient characteristics of the centrifugal pump were tested under various suction pressure and starting conditions. In transient operation of continuous starting and stopping process, instantaneous rotational speed, head, flow rate and suction pressure of the pump were measured. The effect of cavitation on transient performance of the centrifugal pump during transient operation was analyzed, and then the effects of starting acceleration rate and suction pressure of pump on cavitation were presented. Results showed that the cavitation would be delayed during rapid starting period. However, in the condition of low suction pressure and high rotational speed, pump cavitation is inescapable even if the starting period is less than a second. After the serious transient cavitation occurred, the transient performance of centrifugal pump would decline obviously, and the instantaneous head of pump would fluctuate

[en] This work takes the traditional moderator material H₂O for example, the differences between ENDF/B-VIII. β and ENDF/B-VII.1 versions from the ENDF/B nuclear evaluated library are analyzed. In order to compare the thermal neutron scattering data between ENDF/B-VIII. β and ENDF/B-VII.1 versions under the same temperature, the interpolation method is used to obtain the thermal neutron scattering data under arbitrary temperatures. There are some differences between ENDF/B-VIII. β and ENDF/B-VII.1 versions, the thermal neutron scattering cross sections of H bound in H₂O within ENDF/B-VIII. β is different with that of ENDF/B-VII.1, the scattering angle cosines of the neutron and target nuclei from the two libraries are slightly smaller than zero in lower energy range within thermal range. Moreover, the second scattering energy spectrum from ENDF/B-VIII. β is slightly harder than that from the ENDF/B-VII.1 in higher energy range within thermal range when the temperature is higher than room temperature. (authors)

[en] In epithermal energy range, the thermal motion of the target nuclides impacts the elastic scattering. The traditional constant scattering cross section approximation free gas model was adopted in Monte Carlo code to describe the scattering in epithermal energy range. Recent researches show that the bias introduced by ignoring the resonance elastic scattering increases with the temperature. The working temperature of the pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR) is high, therefore it's necessary to consider the resonance elastic scattering by using the Doppler broadening rejection correction (DBRC) method. In this work, the modified Monte Carlo code MCNP 5 was used to analyze the pebble unit cell of PB-FHR. It is found that the DBRC method results in an increase of capture reaction rate of ²³⁸U and a decrease of infinite multiplication coefficient (k_inf) by 123-1182 pcm. In addition, the absolute difference of k_inf introduced by the DBRC increases with TRISO packing factor and temperature. (authors)

[en] Based on the recently released ENDF/B-VII. 1 library, high temperature neutron cross-section files are generated through SIGACE code using low temperature ACE format files. To verify the processed ACE file of SIGACE, benchmark calculations are performed in this paper. The calculated results of selected ICT, standard CANDU assembly, LWR Doppler coefficient and SEFOR benchmarks are well conformed with reference value, which indicates that high temperature ACE files processed by SIGACE can be used in related neutronics calculations. (authors)

[en] In order to develop the temperature-dependent point-wise cross section library for the advanced supercritical water cooled fast reactor, the JEZEBEL fast neutron benchmark was used to analyze the important parameters of the NJOY code and compare the different effects of the input parameters. Then the most reasonable parameters were selected to develop the ASCFR1.0/MC which was based on the ENDF/B-VII.1. Finally, the Doppler coefficient benchmark was applied to test and verify the ASCFR.10/MC. In conclusion, the precision of the ASCFR1.0/MC were perfect. The resulted library can be used in the analysis and verification of the temperature effect in the Advanced Supercritical Fast Reactor (ASCFR), Finally, the moderator effect of ASCFR was calculated with MCNP using the ASCFR1.0/MC library, and the moderator effect of the ASCFR is positive. (authors)

[en] The adoption of inert matrix fuel (IMF) can effectively reduce the amount of minor actinide (MA) in spent fuel of LWR from the source. In order to study the burnup characteristics of IMF, the two typical IMF schemes of PuO_2 + ZrO_2 + MgO and PuO_2 + ThO_2 were selected, and the IMF reactivity with different amounts of PuO_2 was calculated. And the comparison among IMF, UO_2 and MOX was made. The results show that after 36 months operating, with the PuO_2 volume fractions of two types of IMFs from 2% to 10%, their k_i_n_f values at the end of cycle are all above 1. But the burnup reactivity of PuO_2 + ZrO_2 + MgO scheme is greater than that of PuO_2 + ThO_2. The amount of MA at the end of cycle in PuO_2 + ThO_2 scheme is obviously smaller than the former. Under the same effective heavy metal weight percentage, the amounts of MA in MOX and UO_2 fuels at the end of cycle are both greater than those of the two types of IMFs. (authors)

[en] The reactor pressure vessel (RPV) suffers from serious pressurized thermal shock (PTS) which is caused by the cold safety injection fluid direct into RPV downcomer in AP1000. In this paper, the Transient States of Temperature of RPV direct vessel injection nozzle(DVI-Nozzle) in AP1000 under the different operating conditions has been investigated using computational fluid dynamics (CFD) code ANSYS-CFX. Results show that the DVI-Nozzle experiences a large step change in temperature in each operating condition which causes the strong effect on the safety of RPV. (authors)