核动力工程

Preliminary Research of Pebble Bed High-Temperature Gas-Cooled Reactor with Random Refueling Approach Based on VSOP

Wu Hongwei, Xia Bing, She Ding, Li Fu, Zhang Zuoyi

2024, 45(5): 1-6. doi: 10.13832/j.jnpe.2024.05.0001

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Abstract:
The pebble bed high-temperature gas-cooled reactor (PB-HTGR) is characterized by continuous online refueling. The fuel ball flows slowly in the core of the reactor. The VSOP code, which is widely used in engineering design, employs an approximate and average refueling method to simulate the process, tending to diminish the randomness of the pebble flow of PBR to some extent. In this paper, a new approach of random refueling based on VSOP code is proposed, and it improves the core refueling model and focuses on the impact of the average merging effect of discharging fuel. The results show that the random refueling approach can provide a more refined discharge fuel burnup probability distribution, and the average merging effect of discharging fuel tends to the broadening and overlapping of burnup peaks.

Research on Solving Different Nuclear Reactor Models by Coupling Multiphysics Environment (COME) Based on Operator Spliting, Picard and JFNK Methods

Zhou Xiafeng, Zhong Changming, Zhang Yangyi, Zhang Yunshan, Zeng Wei, Tang Qifen, Qiang Shenglong, Gong Zhaohu

2024, 45(5): 7-18. doi: 10.13832/j.jnpe.2024.05.0007

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The research on multi-physics and multi-scale coupling of nuclear reactor is challenging and prominent in the field of nuclear energy. Especially for the large-scale, multi-dimensional and strongly nonlinear coupling system with drastic changes in physical parameters such as temperature, power and density and complex coupling conditions, the current coupling calculation codes often have problems such as slow convergence or even non-convergence, which brings many challenges and difficulties to the development of next-generation coupling computational codes. In recent years, the Virtual Reactor Coupling Analysis Team (VRCAT) at Huazhong University of Science and Technology has developed a unified coupling computational framework called Coupling Multiphysics Environment (COME) based on various coupling methods, including operator splitting, Picard iteration, and Jacobian-free Newton-Krylov (JFNK). This paper first provides a detailed analysis of the main features of COME, including the coupling methods, overall framework, and common interface. Then, based on COME, several coupling problems such as neutron transport/diffusion model, core thermal sub-channel coupling model, two-phase flow coupling model within the system analysis code and complex physical-thermal coupling model are solved respectively, and the convergence and computational efficiency of different coupling methods are compared, so as to provide method guidance and development suggestions for improving the computational stability and convergence characteristics of real complex multi-physical coupling codes.

Development of Computing Code for Full Spectrum Assembly

Zhang Jinchao, Zhang Qian, Zhao Qiang, Zou Hang, Yu Jialei, Wu Shifu, Chen Ying

2024, 45(5): 19-25. doi: 10.13832/j.jnpe.2024.05.0019

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In order to address the issues caused by the diverse geometric designs and complex energy spectrum problems introduced by the use of moderator materials in advanced assembly designs, and to further enhance the design capability of the assembly code, this paper designs a combination strategy and develops a corresponding code based on unstructured mesh. The strategy combines the subgroup method, which is based on a group structure including 2164 groups, with the method of characteristic transport. To ensure computational efficiency, the code employs the efficient multi-nuclide resonance interference method, offset algorithm for scattering source and parallel scheme of graphics processing unit (GPU) characteristics at thousand-group level. Advanced assembly designs with different energy spectra and geometric features are selected to validate the proposed method. The results indicate that, compared to the Monte Carlo reference, the deviations in eigenvalue are within 72pcm (1pcm=10⁻⁵) for problems with fast spectra, and within 132pcm for problems utilizing moderators. In conclusion, the calculation scheme proposed in this paper can handle assembly problems characterized by complex geometry and complex energy spectrum.

Application of Unstructured Mesh Variational Nodal Method in He-Xe Cooled Micro Reactor

Sun Qizheng, Liu Xiaojing, Zhang Tengfei

2024, 45(5): 26-31. doi: 10.13832/j.jnpe.2024.05.0026

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Advanced designs of micro reactors are distinguished by intricate geometries and pronounced neutron leakage. To enhance the precision of neutron property analysis in these complex geometrical designs, this study introduces an Unstructured Mesh Variational Nodal Method (UVNM-S_N). The UVNM-S_N method initiates from the functional formulation of the second-order even-parity neutron transport equation. Spatially, arbitrary triangular unstructured meshes and coordinate mapping techniques are employed. Besides, angular decoupling of the original equations is achieved through the application of the discrete-ordinate (S_N) method. Furthermore, the SIMONS design, a helium-xenon (He-Xe)-cooled micro reactor, is chosen as the analytical subject to verify the performance of the UVNM-S_N. Numerical results demonstrate that UVNM-S_N exhibits geometric adaptability and computational precision in heterogeneous problems with complex geometries. Consequently, it is concluded that UVNM-S_N can stand as an innovative strategy for numerical simulations in advanced micro-reactor designs.

Study on Reactivity Control of Long-life Small Lead-cooled Fast Reactor

Jin Xin, Wang Lipeng, Guo Hui, Chen Lixin, Jiang Xinbiao, Gu Hanyang

2024, 45(5): 32-39. doi: 10.13832/j.jnpe.2024.05.0032

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An effective reactivity control method is needed to compensate for the large reactivity swing during the long lifetime of small lead-cooled fast reactors. However, the compact core structure constrains the arrangement of the control rod system, and the control rod with high reactivity worth can bring safety problems to the core. In this study, the reactivity control of LFR-180 (lifetime reactivity swing is 6681pcm, 1pcm=10⁻⁵), a self-designed small lead-based fast reactor, is studied, the scheme of reactivity control based on flammable poisons is explored, and the transient safety characteristics of accidents are analyzed. The results show that the reactivity swing of LFR-180 can be reduced to 575pcm by using ZrH_1.6 as moderator and B₄C as burnable poison. At the same time, the reactivity control scheme significantly improves the temperature safety margin of the core under the accident of control rod withdrawal (CRW), and ensures the safety of the core under the accident of unprotected loss of primary flow (ULOF) and unprotected loss of heat sink (ULOHS). The burnable poison control scheme established in this study can be applied to the reactivity compensation of small lead-cooled fast reactors with long life.

Research on Monitoring and Control Technology for Axial Xenon Oscillation in Pressurized Water Reactors

Fei Jingran, Bi Guangwen, Yang Bo, Yang Weiyan, Han Yu, Tang Chuntao

2024, 45(5): 40-44. doi: 10.13832/j.jnpe.2024.05.0040

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When the control rod system of the third generation passive advanced PWR (CAP1000/CAP1400) is in manual control mode, it is necessary to provide real-time, accurate and easy-to-read information about xenon distribution in the core to support its rational decision-making and proper operation. Based on the neutron physics model of CAP1000 reactor, this paper studies the monitoring method of xenon oscillation, and develops the function of xenon mode graph in the on-line monitoring system of the third generation passive advanced pressurized water reactor core, which can effectively monitor and control the development of axial xenon oscillation. This scheme has certain applicability and can be extended to other types of PWR plants.

Study on Treatment Method of Environmental Effect of Hexagonal Assembly PWR

Zhang Cheng, Wan Chenghui

2024, 45(5): 45-52. doi: 10.13832/j.jnpe.2024.05.0045

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An improved multi-assembly homogenization method is raised to deal with the environmental effect of the hexagonal fuel assembly in PWRs. In order to reduce the influence of environmental effect of fuel assembly on the calculation accuracy of two-step method, sensitivity analysis is carried out from the state variable of burnup depth, which expands the practicability of this method. In this method, the real core energy spectrum of the fuel assembly adjacent to the reflector is approximately obtained by establishing a multi-assembly model, and the heterogeneous correction factor is adopted to reduce the deviation of the calculation of few-group constants of the single-assembly model with reflective boundary due to environmental effects. At the same time, the traditional two-step calculation strategy is fine-tuned, which has little impact on the overall code framework. The calculation results show that this method can effectively improve the calculation accuracy of the traditional two-step method, and the eigenvalue deviation is reduced from −341pcm (1pcm=10⁻⁵) to −111pcm, and the root mean square deviation of assembly power is also reduced from 2.28% to 1.38%.

Method Research on Neutron-diffusion Solution Based on Arbitrary Quadrilateral Mesh and Conformal Mapping

Guo Lin, Wan Chenghui, Wu Hongchun

2024, 45(5): 53-61. doi: 10.13832/j.jnpe.2024.05.0053

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In order to solve the problem that the traditional "horizontal and vertical" nodal grid is difficult to deal with the deformation of nodal grid caused by the bowing of PWR fuel assembly, this paper studies the neutron diffusion solution method based on arbitrary quadrilateral mesh and conformal mapping according to the nonlinear iterative neutron diffusion solution idea. Arbitrary quadrilateral mesh is used to characterize the deformation of nodal grid caused by fuel assembly bowing, and the global coarse mesh difference finite equation based on arbitrary quadrilateral mesh is established. The arbitrary quadrilateral is transformed into rectangle by conformal mapping, and a local two-node expansion equation based on conformal mapping is established. The numerical results of bowing cases based on two-dimensional mini-core and HPR1000 core indicate that, the core effective multiplication factor (k_eff) and power distribution calculated by the proposed method are in good agreement with the reference NECP-MCX. Therefore, the method proposed in this paper could accurately characterize and simulate the fuel-assembly bowing.

Generation of the Few-Group Cross Sections for Molten Salt Reactors Based on Non-Uniform Spectra Modification Method

Dai Ming, Zhang Ao, Cheng Maosong

2024, 45(5): 62-70. doi: 10.13832/j.jnpe.2024.05.0062

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In view of the high leakage and strong heterogeneity of molten salt reactors, the whole-core transport calculation based on the method of characteristic (MOC) is suitable for the critical calculation of molten salt reactors. To balance computational efficiency and accuracy, the non-uniform spectra modification (nSM) method is proposed to generate the few-group cross sections for MOC whole-core calculations. This method is an embedded leakage modification method, which utilizes the leakage parameters obtained by the few-group MOC whole-core calculation for multi-group spectrum calculations, thereby online updating the few-group cross sections of the whole-core calculation. Compared with results from the continuous energy Monto Carlo code OpenMC for the MSRE two-dimensional whole-core benchmark problems, the k_eff errors calculated by nSM method are less than 0.25% and the maximum error of 2.6% in fission rate distributions occurs at the core can node with a relative fission rate of only 0.04, which are obviously better than those using the cross sections collapsed by the spectra calculated by assembly calculations. The results show that the approximate nSM method still has good accuracy, and could significantly improve the calculation efficiency. Therefore, nSM method could be a feasible method that provides few-group cross sections for the MOC whole-core calculation of molten salt reactors.

Study on Reactivity Change Caused by Dynamic Impact of Space Nuclear Reactor

Wang Lipeng, Cao Lu, Li Rui, Liu Shichang, Chen Lixin, Jiang Duoyu, Hu Tianliang

2024, 45(5): 71-77. doi: 10.13832/j.jnpe.2024.05.0071

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Accurate calculation of reactivity changes caused by impact is a key problem to be solved urgently in the design and safety analysis of micro-reactors. In this paper, neutron transport simulation and explicit finite element dynamics simulation based on unstructured-mesh Monte Carlo are combined to study the multi-physical coupling calculation of micro-reactor under the condition of large deformation of dynamic impact. Taking the 85-pin NaK-cooled space nuclear reactor as an example, the time-dependent variation of reactivity in the process of vertical and 45° dip impact is analyzed. The results show that, without considering the uniform density change of fluid and fuel, the k_eff caused by vertical impact increases by about 8%, while the k_eff caused by 45° impact increases by about 3%, which were in good agreement with the results in the literature. Under the condition of non-uniform fuel density change, the increase of k_eff in the two scenarios increased by about 10% and 20%, respectively. The research provides an important theoretical foundation for the critical safety analysis of the space nuclear reactor’s launch.

Implementation and Optimization of Function Expansion Tallies Based on Track Length Estimation Method in RMC

An Nan, Wang Wu, Wang Kan

2024, 45(5): 78-84. doi: 10.13832/j.jnpe.2024.05.0078

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Traditional Monte Carlo simulations usually use bin-counting to statistically analyze relevant parameters. Rough bin division is difficult to accurately describe the distribution of some parameters in the whole space, while detailed bin division requires a large number of samples to meet the required statistical accuracy, which will take a lot of time. The functional expansion tallies method (FET method) can obtain the continuous distribution of parameters in the solution space, and can solve the problem that computational efficiency and accuracy cannot be achieved at the same time. The FET method based on track-length estimation is innovatively implemented in Monte Carlo Code (RMC). In addition, the Legendre polynomials and Zernike polynomials are combined to calculate the continuous distribution of the parameters in the three-dimensional assembly space. At the same time, the simulation time of FET method and meshtally method are compared. The results show that the calculation results of FET method are in good agreement with the meshtally method, and the simulation time of FET method is reduced while the simulation memory is greatly reduced. Therefore, the functional expansion tallies method developed in this study can be used in Monte Carlo code.

Study on Two-Phase Flow Instability in Helical Tube under Rolling Condition

Xian Lin, Cheng Kun, Ran Xu, Wu Dan, Yan Junjie, Qiao Shouxu

2024, 45(5): 85-91. doi: 10.13832/j.jnpe.2024.05.0085

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Spiral-tube steam generator has the advantages of compact structure and strong heat exchange capacity, and it becomes increasingly prevalent in reactor design. However, its flow and heat transfer characteristics are different from those of straight-tube heat exchanges under marine conditions. Particularly, the instability of two-phase flow under rolling condition remains inadequately explored. In this study, the experimental study on the two-phase flow instability of a single helical tube is carried out under static and rolling conditions, and the process of its transition from single-phase flow to density wave pulsation and then to pressure drop pulsation under different heating power levels is studied. Under the static condition, when the heating power is low, the fluctuation range of each parameter of single-phase flow in the helical tube is within 1%. When the heating power reaches 11 kW, the density wave pulsation with a period of 4.4s is generated, and when the heating power reaches 13 kW, the pressure drop pulsation with a period of about 34.3s is generated. Under the rolling condition, the rolling motion and pulsation have a significant compound effect, and the fluctuation period and amplitude have changed. By studying and processing the experimental data, the characteristics of the period and frequency of the two-phase flow instability in the helical tube are obtained, and the mechanism that causes the difference between the two-phase flow instability in the helical tube and the straight tube flow channel is revealed, as well as the influence mechanism of the rolling condition on the two-phase flow instability.

Analysis of Convective Heat Transfer Characteristics and Entropy Generation of Fluid in Fish-scale Bionic Enhanced Heat Transfer Tubes

Liu Xiaoya, Zhao Xinwen, Xiao Hongguang, Ran Lingke, Zhang Yinxing, Zhang Yongfa, Sun Jichen, Ding Ming

2024, 45(5): 92-98. doi: 10.13832/j.jnpe.2024.05.0092

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With the development of bionics, bionic technology has better drag reduction and heat transfer effects. Inspired by the fish-scale bionic concept, three biomimetic enhanced heat transfer tubes are proposed. The effects of different depths, spacings and angles of bionic enhanced heat transfer tubes on the fluid flow and heat transfer characteristics in the turbulent flow regime (Re=15700−62900) are studied by numerical simulation. The results show that the three bionic enhanced heat transfer tubes all have perfect heat transfer enhancement effect. The greater the depth, the smaller the spacing and the smaller the angle, the better heat transfer enhancement effect. Under the same conditions, the comprehensive performance and entropy generation of bionic enhanced heat transfer tubes are analyzed, and it is found that the second type has the best comprehensive performance, the largest performance evaluation criteria (PEC) and the smallest power loss.

Preliminary Analytical Study of the Effect of Accident Tolerant Fuel on Fuel Rod Performance under LOCA Condition

Wang Zeji, Guo Zhangpeng, Zhu Aobo, Ouyang Xiaoping, Niu Fenglei

2024, 45(5): 99-107. doi: 10.13832/j.jnpe.2024.05.0099

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Accident tolerant fuel (ATF) cladding material is a new generation of nuclear fuel concept proposed after the Fukushima nuclear accident to improve the performance of fuel elements against severe accidents. Compared with the current Zr-4 alloy cladding material, ATF cladding material can resist the consequences of accidents for a long time, while maintaining or improving its performance under normal operating conditions. In this paper, based on the code FRAPTRAN-2.0 and two ATF cladding materials (FeCrAl and SiC), a transient analysis code of fuel rod performance for ATF cladding materials was developed by improving the thermalphysical model, mechanical behavior model and oxidation model of cladding materials. Then the fuel rods of the MT-1 experimental stand were used as the objects for the computational analysis of its LOCA condition, and the thermal-hydraulic transient response characteristics of ATF cladding materials under this condition are studied. The analysis results show that compared with the conventional Zr-4 alloy cladding, the ATF cladding material can not only reduce the peak cladding temperature (PCT) under LOCA, but also delay or prevent cladding failure.

Analysis of Influence of Expansion on Control Rod Drop in Nuclear Reactor at High Temperature

Chen Changyi, Xi Yanyan, Lu Yaheng, Wu Xuanlong, Wu Feng

2024, 45(5): 108-114. doi: 10.13832/j.jnpe.2024.05.0108

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The high temperature environment in the nuclear reactor will cause the thermal expansion of the control rod assembly, which will affect the falling time of the control rod assembly. In order to study the effect of thermal expansion on the drop of control rod assemblies, a fluid-structure coupling model for the drop of control rod assemblies was established considering expansion deformation, and the dynamic equations for the drop process of control rod assemblies were solved. The steady state thermodynamic coupling analysis of the guide tube and control rod is performed using finite element method to obtain the expansion deformation of the control rod and guide tube under high temperature conditions. Based on the established analytical model, the dynamic processes of control rod drop with and without expansion is compared. The analysis results show that the thermal expansion phenomenon delays the total drop time of the control rod, but has little impact on the drop time before the buffer section. Therefore, in general, the impact of thermal expansion on the drop time can be ignored, but the consequences caused by the control rod expansion still need to be paid attention to in the engineering design. The conclusion of this paper is important for the design of control rod and the analysis of in-pile control rod drop time.

Preliminary Engineering Validation of the High-fidelity Multi-Physical Coupling Code CRANE/EAGLE

Chen Guohua, Feng Jinjun, Chen Chao, Jiang Xiaofeng, Wang Tao

2024, 45(5): 115-120. doi: 10.13832/j.jnpe.2024.05.0115

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Based on the numerical reactor physical code CRANE and sub-channel thermal hydraulic code EAGLE accelerated by GPU, a set of CRANE/EAGLE high fidelity multi-physical coupling software system was developed through direct joint compilation at the source code level. A large amount of verification and validation work has been carried out on the CRANE/EAGLE system. This paper mainly introduces the verification results of the first cycle of Tianwan Unit 5 (M310 reactor) and Unit 4 (VVER-1000 reactor). According to the first cycle of these two units, the calculation of starting physical parameters and the tracking simulation of operation history in days are carried out. The numerical results show that the CRANE/EAGLE code system not only has high calculation accuracy, but also can complete the multi-physics coupling calculation of a single state point of a commercial pressurized water reactor in minutes on a small multi-GPU computing platform. The CRANE/EAGLE software system verified in this paper has preliminary engineering application value.

Numerical Simulation of the Natural Circulation Test of PHENIX Reactor by ACENA

Liu Yapeng, Zhang Dalin, Chen Yutong, Zhou Lei, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(5): 121-127. doi: 10.13832/j.jnpe.2024.05.0121

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Natural circulation decay heat removal is is an important passive safety feature of pool-type sodium-cooled fast reactor (SFR), but it also brings new problems and challenges to the design and safety analysis of SFR. Whether the natural cycle can be established and whether the natural cycle can take away the decay heat of the core is an important part of the reactor safety analysis under station black-out accident (SBO). In order to verify the reliability of the liquid metal accident analysis code ACENA in the simulation of natural cycle decay heat removal in SFR, the natural circulation test conducted by the French Alternative Energies and Atomic Energy Commission (CEA) at the end of the life of pool-type sodium-cooled fast reactor PHENIX was modeled and analyzed. The effect of thermal stratification on natural circulation was considered by two-dimensional finite difference method in sodium pool. The verification results show that ACENA can accurately analyze the process of forced circulation to natural circulation and accurately predict the key parameters. The code can also calculate the thermal stratification phenomenon in the sodium pool, and has the calculation capacity of natural circulation decay heat removal.

Research on Heat Transfer Characteristics of Corium Pool Under Oscillating Conditions

Luo Simin, Zhan Dekui, Chen Peng

2024, 45(5): 128-135. doi: 10.13832/j.jnpe.2024.05.0128

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In order to obtain the heat transfer characteristics of corium pool under oscillating conditions, an experimental study was carried out on the two-layer corium pool of small marine reactors. Fluorinert liquid FC-40 and water were used respectively to simulate the oxide layer and metal layer of corium pool in the experiment, and the transient variations of temperature field and heat transfer capacity under different oscillating conditions were obtained. The experimental results show that the oscillating conditions exert the most intense influences in the early stage of movements, and as the oscillating motion continues, the corium pool will reach a quasi-steady state of thermal equilibrium. In general, the temperature stratification of corium pool is weakened under the condition of oscillation, and the overall temperature is lower than that under the static condition, together with the increased heat transfer capacity to the cooling wall. Under the same high-intensity oscillating conditions, the effect of oscillations in vertical direction is more intense than that of oscillations in lateral direction. However, under the condition of low-intensity oscillation, the effect of oscillations in vertical direction can be ignored. Additionally, a new dimensionless parameter Lo was proposed to characterize the oscillatory influence strength, which represents the ratio of the characteristic oscillatory force to the characteristic shear force of fluid under oscillating conditions. It can be used to quantify the influence of different oscillation intensities under the same oscillation direction. This research is supposed to offer valuable reference to the in-vessel retention (IVR) analysis and safety system design for small marine reactors.

Scaling Experimental Design of Horizontal Steam Generator in NHR200-II

Li Zongyang, Hao Wentao, Zhang Wenwen, Li Weihua, Yang Xingtuan

2024, 45(5): 136-141. doi: 10.13832/j.jnpe.2024.05.0136

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To broaden the application range and improve the economic feasibility of 200 MW Low-Temperature Nuclear Heating Reactor (NHR200-Ⅱ), a redesign of its intermediate loop has been proposed, including a new superheated steam supply system that incorporates horizontal steam generators (HSG). Due to the high cost and long period of prototype scale verification tests, a scaling modeling analysis of the two-phase natural circulation within the HSG was employed using the Hierarchical Two Tiered Scaling (H2TS) method. Theoretical derivations provided scaling analysis criteria under isothermal conditions and scaling numerical values under various scaling conditions. Both the prototype and the model employed water as the test fluid, and numerical modeling of the prototype and the model was conducted by using the RELAP5 code. The numerical calculations indicated that the key parameters’ proportional relationships between the model and the prototype were consistent with the theoretically derived results. Considering factors such as test economy, accuracy, and safety, a 1:4 scale model-to-prototype length ratio was ultimately selected as the scaling factor for subsequent verification tests, corresponding to a power ratio of 1:96.

Study on Fretting Wear Behavior of Pre-oxidized Zircaloy Cladding in High Temperature and High Pressure Water

Wang Jun, Wang zhiguo, Cai Zhenbing, Li Zhengyang, Ren Quanyao, Liu Xiaohong, Jiao Yongjun

2024, 45(5): 142-154. doi: 10.13832/j.jnpe.2024.05.0142

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To further study the fretting wear of of claddings with the change of oxidation time in practical service, a variety of pre-oxidized claddings were prepared by superheated steam oxidation. In this study, a self-made high-temperature and high-pressure tangential fretting wear tester was used to carry out fretting wear tests simulating the operation conditions of PWR, and the volume wear coefficients of the substrate and the cladding after pre-oxidation at different times were measured. The results show that the surface hardness of the cladding is 2~3 times higher than that of the substrate, and the wear coefficient is reduced by about 90%. A dense oxide layer formed on the surface layer of cladding is an important reason for the change in its wear coefficient. The longer the oxidation time, the thicker the oxide layer, and the cladding with an oxidation time of 200 d has the lowest wear coefficient. In addition, the existence of oxide layer causes the fretting wear mechanism of zircaloy cladding to change from serious abrasive wear and layering to slight abrasive wear and adhesive wear in high temperature and high pressure water environment.

Research Progress and Technological Development Trend of Accident Tolerant Fuel

Li Ziyi, Wang Xiaomin, Wang Kai, Zhang Ruiqian, Yin Chunyu, Chen Huan, Shi Haojiang, Pei Jingyuan, Lu Yonghong

2024, 45(5): 155-164. doi: 10.13832/j.jnpe.2024.05.0155

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The research and development of accident tolerant fuel (ATF) has become a new research direction in the international fuel industry in the post-Fukushima era, which involves the research and development of advanced cladding materials and new nuclear fuels. After more than ten years of comprehensive and systematic research, the international nuclear fuel industry represented by the U.S. and France have gained important progress, focusing further on medium and long-term technical solutions. In this paper, the important progress, challenges and development trend of subsequent technologies in ATF cladding materials (including Cr coating, FeCrAl alloy and SiC composite) and fuels (including enhanced UO₂, high uranium density fuel and ceramic dispersive matrix dispersion (CDM) fuel) at home and abroad are reviewed.

Development of Segmented Air-gap Coupled Electric Heating Test Device for Material Irradiation

Huang Gang, Si Junping, Sun Sheng, Jin Shuai, Peng Xingjie, Tong Mingyan, Peng Fang

2024, 45(5): 165-170. doi: 10.13832/j.jnpe.2024.05.0165

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In order to solve the technical problem of precise control for the material irradiation temperature in high heat release rate region of the research reactor, the segmented air-gap coupled electric heating test device for the material irradiation was developed. The irradiation temperature control method of axial segmented air gap coupling electric heating was adopted to enable the irradiation device to have a large temperature regulation ability. The structure of axial segmented yin-yang plane independent partition was adopted in the test device. The irradiation test section loaded with material samples was composed of four independent chambers, each of which was equipped with an independent inert gas regulating circuit, and the two chambers on the first segment were equipped with the electric heating rod auxiliary air gap for the irradiation temperature regulation. This device was used to carry out irradiation test in the research reactor, and the results show that the material irradiation temperature for the test device can be effectively controlled at 335~365℃ in high flux rate region of the research reactor, and the independent temperature control ability of the electric heating rod can reach 30℃. The developed device greatly improves the accuracy of irradiation temperature control, realizes on-line accurate adjustment and control of irradiation temperature, and achieves the purpose of fine temperature control for material irradiation.

Study on Added Mass and Fluid Damping Characteristics Based on Concentric Cylindrical Structure

Zhu Shibin, Ai Huaning

2024, 45(5): 171-176. doi: 10.13832/j.jnpe.2024.05.0171

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To deeply investigate the intrinsic characteristics of added mass and fluid damping, analyze the impact of viscosity and amplitude on them, and provide guidance for analyzing fluid-induced vibration, this study takes a concentric cylinder as an example to establish a prediction method of added mass and fluid damping based on computational fluid dynamics (CFD). The user-defined function (UDF) is used to set the motion equation of the inner circle, and the overset grid technology is used to complete the grid motion, so as to realize the numerical simulation of the flow field. The shape of the function is determined according to Bearman's hypothesis, and the calculated fluid force curve is fitted by the least square method to obtain the added mass and fluid damping. Finally, the influences of viscosity and dimensionless amplitude on results are compared. The calculation and analysis results show that the viscosity not only affects the fluid damping but also the added mass. The dimensionless amplitude has little effect on the added mass and an obvious effect on the fluid damping. Pressure damping and viscous damping increase in equal proportion with the increase of dimensionless amplitude, and the proportion of pressure damping increases with the decrease of diameter ratio. The solution of the modified formula with dimensionless amplitude effect is in good agreement with the numerical results. The research in this paper has an important guiding role in optimizing the existing analysis methods of flow-induced vibration.

Research on PSA Event Tree of Excessive Radioactive Release Risk for Aqueous Homogeneous Reactor

Zou Zhiqiang, Zhang Dan, Liu Yu, Wang Ningning, Sun Hongping, Wang Zhe, Yang Weidong, Du Yu

2024, 45(5): 177-183. doi: 10.13832/j.jnpe.2024.05.0177

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The design characteristics and safety features of aqueous homogeneous prototype medical isotope research reactor (AHR) are significantly different from those of traditional solid fuel pressurized water reactor, thus the purpose and range of probabilistic safety analysis (PSA) of these two kind of reactors, especially the event tree analysis are also different. Taking the AHR as the research object, the differences between its fuel form, safety barrier and mitigation system and that of the traditional PWR are analyzed. The radioactive release path, containment boundary and the main accident types that should be considered in the internal event PSA for the purpose of analysis of excessive radioactive release are determined, and the universal excessive radioactive release event trees of typical accident categories are constructed. The study can provide guidance for the PSA and further risk quantification of radioactive release risk of this type of reactor.

Research on Nuclear Power System Modeling and Control Methods Based on APROS

Xie Aoda, Yang Ting

2024, 45(5): 184-191. doi: 10.13832/j.jnpe.2024.05.0184

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As nuclear power system modeling increasingly moves towards greater accuracy and refinement, research on the coupling of three-dimensional core and thermal-hydraulic models has also grown, which provides a better model foundation for the design of control systems. In this study, the APROS software was used to model the coupling of a three-dimensional core and thermal hydraulics for VVER-1000. A load-following control system based on model predictive control (MPC) and other control systems were designed. The model was then verified using steady-state and transient simulation results, which showed that the model simulated effectively. Taking advantage of the three-dimensional core visualization, the performance and safety of the MPC load-following controller were further validated. This research not only provides a model foundation for studies on nuclear power systems but also offers practical experience for the safety analysis of advanced power control systems.

Analysis of Station Blackout Accident in Nuclear Power Plant Based on Dynamic Event tree

Chen Haoyin, Wang He, Zhao Qiang, Li Lei, Wang Longcong

2024, 45(5): 192-198. doi: 10.13832/j.jnpe.2024.05.0192

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Abstract:
Traditional event tree analysis ignores the dynamic time parameters and heavily relies on expert judgement. To tackle these issues, this study utilizes dynamic event tree (DET) method to to establish a response model for station blackout (SBO) accidents of China's improved three-loop PWR (CPR1000) nuclear power plant. Time-dependent branching is established for branch nodes such as turbine-driven auxiliary feedwater system and AC power recovery, and the accident branching results are comprehensively simulated, and the branching probability and the failure probability of core damage under SBO accident are calculated. The calculation results show that different running time of turbine-driven auxiliary feedwater system and recovery time of AC power supply have obvious influence on the accident consequences. The increase of running time of turbine-driven auxiliary feedwater system can prolong the time window of power supply recovery, but there is an upper limit for power supply recovery time, beyond which core damage is inevitable. Compared with the failure probability calculated by the traditional event tree, the dynamic event tree method reduces the failure probability value and develops the potential safety margin.

Temperature Rise Analysis of Control Rod Drive Mechanism Based on Equivalent Thermal Network Method

Xu Qiwei, Liu Sheng, Luo Lingyan, Yu Tianda, Fu Guozhong, Yang Yun, Zhao Yizhou

2024, 45(5): 199-205. doi: 10.13832/j.jnpe.2024.05.0199

Abstract(29) HTML (11) PDF(10)

Abstract:
The Control Rod Drive Mechanism (CRDM) in nuclear reactor operates in the harsh environment of high temperature, high pressure, and high radiation for a long time. In order to effectively prevent the damage caused by excessive temperature, it is necessary to predict and estimate the temperature of internal components of CRDM to improve the safety and reliability of nuclear reactor. This paper proposes a temperature estimation method based on the equivalent thermal network method. First, the heat source parameters required for the temperature rise analysis are calculated. Then, the thermal resistances in different regions are solved to construct an equivalent thermal network model of the CRDM, enabling rapid and accurate estimation of the CRDM temperature. Finally, the results are validated using Finite Element Analysis (FEA). The validation results show that the proposed CRDM temperature estimation method has high accuracy, providing a theoretical basis for subsequent novel structural design and reliability analysis.

Safety Management Strategies and Practices for Multi-Unit Group Operation at Tianwan Nuclear Power Plant

Yao Gang

2024, 45(5): 206-212. doi: 10.13832/j.jnpe.2024.05.0206

Abstract(42) HTML (12) PDF(9)

Abstract:
At present, the nuclear power units in operation and under construction in Tianwan Nuclear Power Plant (TNPP) include three different types of units, namely, VVER-1000 (Units 1-4), M310 (Units 5 and 6) and VVER-1200 (Units 7 and 8), covering a variety of nuclear power design technical directions. Under the multi-unit group operation mode, TNPP has carried out research on the optimization of operatng safety management strategy based on the design characteristics of different units, and established a unified Symptom-oriented Emergency Operating Procedure (SEOP) system for nuclear power units, which has the characteristics of accurate and rapid Event-oriented Procedure (EOP) and wide coverage of State-oriented Emergency Operating Procedure (SOP). At the same time, the technical framework of the new version of Severe Accident Management Guideline (SAMG) implemented by Units 7 and 8 contains more comprehensive ability to deal with serious accidents and simpler diagnostic guidelines, which can effectively alleviate the consequences of serious accidents in nuclear power plants. The new SAMG technical framework can also be applied to other units of TNPP.

Study on Stability Criteria for Leakage Rate Measurement of Non-adiabatic Containment

Li Jianfa, Liu Mingmei, Hua Yongzhen, Chu Weiyu, Sun yifan

2024, 45(5): 213-218. doi: 10.13832/j.jnpe.2024.05.0213

Abstract(51) HTML (12) PDF(9)

Abstract:
Current standards such as ANSI/ANS56.8 and NB/T20018—2021 require the assumption of containment insulation for measuring the leakage rate of containment. The gas quality stability criteria in the standards will fail in the measurement of the leakage rate of non-adiabatic containment. In order to explore the sealing evaluation method of non-adiabatic containment, this study innovatively proposed a leakage rate stability criterion and completed experimental verification on the non-adiabatic scientific research containment at Langfang R&D Base of China Nuclear Power Engineering Co., Ltd. The results indicate that the new stability criterion for leakage rate can be used to measure the leakage rate of non-adiabatic containment. In addition, the test data of a certain nuclear power plant indicates that the new criterion can also be used for measuring the leakage rate of adiabatic containment, shortening the test time. The conclusion of this study can support the application of scientific research non-adiabatic containment to conduct leakage rate research and optimize the sealing test technology of nuclear power plant containment.

Common Faults and Solutions in the Process of Electric Valve Commissioning of EPR Nuclear Power Unit

Shen Kun, Guo Shusheng, Li Derui, Fang Xing

2024, 45(5): 219-224. doi: 10.13832/j.jnpe.2024.05.0219

Abstract(27) HTML (8) PDF(10)

Abstract:
There are 926 electric valves in an evolutionary power reactor (EPR) nuclear power plant in China. During the commissioning of electric valves, six common faults were found in the electric actuators, namely, deviation between the torque value and the designed value, frequent triggering of the open torque microswitch, reversal, lack of jumper in the control loop, unreasonable selection of gear ratio in the reduction gear, and leakage of lubricating grease in the motor gearbox. This paper addresses these six common faults by providing detailed descriptions of the fault phenomena and analyses of the fault causes, and then proposes corresponding fault-solving methods, including torque calibration of the electric actuator, adding a bypass for the open torque microswitch, adjusting the phase sequence of the cables, adding a jumper, replacing the gear with the appropriate number of teeth in the reduction gear, and replacing the seal ring. The solution to these six common faults not only improves the reliability of the electric valve operation, but also provides experience for the subsequent design and commissioning of nuclear power unit electric valves.

Research on Underwater Laser Welding System and Welding Process of S32101 Duplex Stainless Steel Cladding in Spent Fuel Pool

Zhang Xiaochun, Shen Guangyao, Mei Le, Zhu Jialei, Li Congwei

2024, 45(5): 225-231. doi: 10.13832/j.jnpe.2024.05.0225

Abstract(72) HTML (19) PDF(9)

Abstract:
In order to achieve high-quality underwater repair of spent fuel pool cladding defects in nuclear power plants, an underwater laser welding system with mobile positioning system, underwater laser-vision-wire feeding integrated equipment and mobile gas hood was developed based on underwater local dry laser welding technology. The underwater laser welding system was used to optimize the underwater welding process of the third generation nuclear spent fuel pool steel cladding (S32101 duplex stainless steel). The results show that the influence of water molecules on weld formation can be effectively reduced by using the drainage technology of equalizing flow pipe and simultaneous purging of groove. Using nitrogen as the protective gas for underwater welding of duplex stainless steel can increase the austenite content in the cladding layer and heat affected zone. The welding defects can be reduced by use of higher laser power, thus improving the stability of underwater local dry laser welding. The results of non-destructive testing and physical and chemical properties of welded specimens meet the requirements of welding standards for spent fuel pool construction, which proves the feasibility and reliability of the underwater laser welding system and welding process.

Research on Phased Array Ultrasonic Inspection Technology for the Pressurizer Surge Nozzle to Vessel Weld

Luo Liqun, Zhu Jiazhen, Chen Jun, Kang Zhiping

2024, 45(5): 232-236. doi: 10.13832/j.jnpe.2024.05.0232

Abstract(28) HTML (7) PDF(11)

Abstract:
In response to the single-sided accessibility issue caused by the weld structure and the significant positioning deviation caused by spherical detection surface in the surge nozzle to vessel weld of pressurizer in nuclear power plant, a phased array ultrasonic inspection (PAUT) technology and a defect positioning correction algorithm are developed, which are verified in a simulated body with natural defects and compared with the results of radiographic inspection. The results show that, combined with the defect positioning correction algorithm, the PAUT technology meets the requirements of the performance demonstration, and the detection capability is equivalent to that of radiographic inspection.

Nuclear Power Measurement System Modification Design for Qinshan Nuclear Power PhaseⅡ Units 1&2

Zhang Yun, Wang Yinli, Tian Ye, Luo Wei, Huang Youjun, Zhuo Xianglin, He Jiaji, Li Mengshu, Sun Qi

2024, 45(5): 237-242. doi: 10.13832/j.jnpe.2024.05.0237

Abstract(63) HTML (14) PDF(14)

Abstract:
Based on the present situation of nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, focusing on the characteristics and existing problems of the original system, this paper analyzes the necessity of the modification of nuclear power measurement system, and introduces the scope of digital modification of nuclear power measurement system. Through the upgrading design of the nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, the design concept, design principle and design flow of the digital modification of the nuclear power measurement system are discussed, and the framework structure design, design characteristics and specific optimization measures of the upgrading of the nuclear power measurement system are given. There were no design changes during the upgrading of the ex-core nuclear measurement system, and the equipment on site was successfully commissioned at the first try and successfully put into operation. The modification scheme and experience can be used as a reference for the change of nuclear measurement system in other nuclear power plants.

Study on Key Factors of ²⁵²Cf Production in Super High Flux Reactor

Xie Yunli, Wang Lianjie, Cai Yun, Xia Bangyang, Huang Xueliang, Lou Lei

2024, 45(5): 243-248. doi: 10.13832/j.jnpe.2024.05.0243

Abstract(32) HTML (10) PDF(12)

Abstract:
²⁵²Cf is widely used in reactor startup and neutron activation analysis, and it is of great significance to carry out research on irradiation production technology. Production of ²⁵²Cf is not easy because of its huge nucleon number and long nuclide transmutation chain. Based on the lead-cooled fast neutron super high flux research reactor, this paper focuses on the key technologies of its irradiation production. According to the difficulty of ²⁵²Cf production, a precise nuclide databank and longtime burnup calculating procedure of nuclide production targets are established, and the key factors such as target structure and material design, neutron energy spectrum and neutron flux density are studied. The irradiation calculation of ²⁵²Cf target shows that although ²⁵²Cf has a low transmutation productivity, it can be improved with appropriate designs of ²⁵²Cf target, neutron energy spectrum and neutron flux density, thus improving the production efficiency of ²⁵²Cf. A method of optimizing neutron local energy spectrum through resonance shielding is proposed to reduce fission consumption of nuclides, thus improving the production efficiency of ²⁵²Cf. This paper expounds the production mechanism of ²⁵²Cf and the influence law of key factors, and gives the demonstration direction of ²⁵²Cf irradiation production technology design.

Application of Orthogonal Experimental Design in Fuel Rod Test Case Design

He Kangnian, Ding Ming, Guo Zehua, Zhu Yanan, Wei Chao

2024, 45(5): 249-255. doi: 10.13832/j.jnpe.2024.05.0249

Abstract(34) HTML (10) PDF(8)

Abstract:
In order to assess the impact brought about by parameter uncertainty on fuel elements performance, combinations of parameters within the error range of the fuel elements at the design stage need to be calculated and evaluated. Orthogonal experimental methods were used to take values for combinations of key parameters of the rod fuel elements and to prepare test cases. Mixed-level orthogonal tables were prepared based on the selected key design parameters and the levels of each parameter, and 36 sets of test cases were identified. Test calculations were performed on the 36 cases, and the results showed that the total cladding deformation of some test cases did not satisfy the fuel rod design guidelines, indicating that the parameter uncertainty negatively affected the fuel element behavior. The preparation and analysis of the test cases provide guidance for the design of fuel elements.

Research on Multi-scale Coupling Simulation of Rod Bundle Channels Based on Subchannel-CFD Coupling Code

Liu Luguo, Jiang Guangming, Xia Yunfeng, Liang Yu, Wang Mingjun, Liu Yu

2024, 45(5): 256-261. doi: 10.13832/j.jnpe.2024.05.0256

Abstract(83) HTML (15) PDF(27)

Abstract:
When the subchannel code is used for core thermal hydraulic analysis, it is necessary to give the real-time information such as the state parameters of the core inlet, and the computational fluid dynamics (CFD) code can calculate the fine thermal hydraulic parameters of the core inlet. In this paper, the subchannel code CORTH is explicitly coupled with the CFD code FLUENT by means of internal coupling method and dynamic link library technology. The coupling code is further used to simulate the experimental conditions of PNL 2×6 benchmark problem, in which FLUENT calculates the inlet section to provide accurate inlet flow distribution for CORTH, while CORTH calculates the simulated heating section. The results show that the multi-scale coupling code can realize the real-time transmission of thermal parameter information between sub-channel code and CFD code, and the simulation results are in good agreement with the experimental results.

Study on Effect of Rolling Motion on Natural Circulation Flow Instability in Rod Bundle Channel

Cheng Kun, Xian Lin, Ran Xu, Zhou Ke, Yu Na

2024, 45(5): 262-268. doi: 10.13832/j.jnpe.2024.05.0262

Abstract(38) HTML (15) PDF(18)

Abstract:
In order to obtain the influence of rolling motion condition on the natural circulation flow stability, the natural circulation loop experimental system with a rod bundle heating channel is used to carry out a comparative experimental study of flow instability under static and rolling conditions. The experimental results show that the flow instability existing in the rod bundle channel is density wave oscillation (DWO). An empirical correlation for the DWO boundary prediction at low pressure NC condition is obtained. Two typical two-phase flow instabilities in rod bundle channel under rolling condition are found in experiments: the trough-type oscillation caused by the vapor generation at the lowest point of flow fluctuation and the compound oscillation formed by the superposition of the trough-type oscillation and DWO. The evolution law of flow instability behavior of natural circulation under the influence of rolling as well as its instability boundary are also obtained.

Research on System Performance and Engineering Application of Stirling Power Conversion Technology

You Ersheng, Zhang Ting, Xing Dianchuan, Xu Jianjun, Yan Xiao

2024, 45(5): 269-276. doi: 10.13832/j.jnpe.2024.05.0269

Abstract(42) HTML (10) PDF(12)

Abstract:
In order to meet the needs of small modular nuclear power plant for the type selection, demonstration and overall evaluation of new power conversion technology, this study focuses on the free piston Stirling power conversion technology and the key technical problems in engineering application, comparing and analyzing the typical application concepts of the hundred-watt isotope-Stirling power supply system, the kilowatt nuclear reactor-Stirling power supply system, and the ten-kilowatt solar power-Stirling power supply system. The performance estimation models of two key parameters, output power and conversion efficiency, of Stirling power conversion system are given. The research results show that the conversion efficiency of Stirling engine is able to reach about 60% of the Carnot cycle, while the output power is generally below tens of kilowatts. For more power generation through the modular configuration of multiple units, the coupling structure and heat transfer process will be the key factors affecting the performance of Stirling engine. The estimation model and analysis results proposed in the paper are useful to support the performance evaluation and engineering application of Stirling power conversion technology.

Current Issue 2024 Vol. 45, No. 5

Current Issue
2024 Vol. 45, No. 5