Performance of different absorber materials and move-in/out strategies for the control rod in small rod-controlled pressurized water reactor: A study based on KLT-40 model

Wu Zhiqiang,Xie Jinsen,Chen Pengyu,Xiao Yingjie,Ni Zining,Liu Tao,Deng Nianbiao,Sun Aikou,Yu Tao 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.7

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638–698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from 0.69 and + 0.81 to 0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).

Selection of burnable poison in plate fuel assembly for small modular marine reactors

Shikun Xu,Tao Yu,Jinsen Xie,Zhulun Li,Yi Xia,Lei Yao 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.4

Small modular reactors have garnered considerable attention in the recent years. Plate fuel elementsexhibit a good application prospect in small modular pressurized water reactors for marine applications. Further, improved economic benefits can be achieved by extending the core lifetime of small modularreactors. However, it is necessary to realize a large initial residual reactivity for achieving a relatively longburnup depth finally. Thus, the selection of a suitable burnable poison (BP) is a crucial factor that shouldbe considered in the design of small modular reactors. In this study, some candidate BPs are selected torealize the effective control of reactivity. The results show that 231Pa2O3, 240Pu2O3, 167Er2O3, PACS-J, andPACS-L are ideal candidates of BP, and since the characteristics of BP can increase the final burnup depthof assembly, the economic benefits are gained. Additionally, an optimal combination scheme of BPs isestablished. Specifically, it is proved that through a reasonable combination of BPs, a low reactivityfluctuation during the lifetime can be achieved, leading to a large final burnup depth

Neutronic Study of Utilization of Discrete Thorium-Uranium Fuel Pins in CANDU-6 Reactor

Nianbiao Deng,Tao Yu,Jinsen Xie,Zhenping Chen,Qin Xie,Pengcheng Zhao,Zijing Liu,Wenjie Zeng 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.2

Targeting at simulating the application of thorium-uranium (TU) fuel in the CANDU-6 reactor, this paperanalyzes the process using the code DRAGON/DONJON where the discrete TU fuel pins are applied in theCANDU-6 reactor under the time-average equilibrium refueling. The results show that the coolant voidreactivity of the assembly analyzed in this paper is lower than that of 37-element bundle cell withnatural uranium and 37-element bundle cell with mixed TU fuel pins; that the max time-averagechannel/bundle power of the core meets the limits - less than 6700kW/860 kW; that the fuel conversionratio is higher than that of the CANDU-6 reactor with natural uranium; and that the exit burnupincreases to 13400 MWd/tU. Thus, the simulation in this paper with the fuel in the 37-element bundlecell using discrete TU fuel pins can be considered to be applied in CANDU-6 reactor with adequatemodifications of the core structure and operating modes.

A fast sampling algorithm for energy-angle distributions of bremsstrahlung photon for radiotherapy applications

Wasaye Muhammad Abdul,Yu Tao,Xie Jinsen,Chen Zhenping,Ni Zining 한국물리학회 2023 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.82 No.10

A fast sampling algorithm for energy-angle distributions of bremsstrahlung photon for radiotherapy purposes is presented. Efcient and accurate sampling methods have been developed based on the most accurate and reliable diferential cross-sections for sampling the energy-angle distributions of bremsstrahlung photons by incident electron of energy 1 MeV–20 MeV. Scaled energy-loss numerical diferential cross-sections produced by Seltzer and Berger are used to sample the photon energy. A new sampling method based on a double diferential cross-section of Koch and Motz has been developed which uses a simplifed expression to sample the bremsstrahlung photon angular distribution. The average efciency of sampling photon energy distribution algorithm is about 85% for electrons with kinetic energies 5 MeV–20 MeV and below to 75% for electrons of kinetic energies 1 MeV–5 MeV. Computation time comparisons have been evaluated with the previous algorithm to sample one photon energy. The inverse transform sampling procedure is implemented to sample the photon angular distribution; hence, every sampled value of the angular distribution is accepted. Therefore, the proposed algorithm is very fast and efcient for radiotherapy purposes. The sampling methods’ accuracy is checked by comparing the Monte Carlo sampled distributions with the theoretical expressions.

Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Zhao, Pengcheng,Liu, Zijing,Yu, Tao,Xie, Jinsen,Chen, Zhenping,Shen, Chong Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.12

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.