bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Preliminary Study on Neutronic Performance of the Spallation Target With the Proton Beam Variation
Authors: Zhou Junjie , Zeng Qin , Yang Jinchen , Yang Yi , Shi Ying , Jiang Yanyi 
Published: October 2021

Abstract
Spallation particles can be utilized in different fields such as neutron scattering studies, external source for burning spent fuel as well as running subcritical reactors. Accelerator driven subcritical system (ADS) has been considered as an advanced nuclear waste transmutation facility with inherent safety feature. In this paper, the neutronic performance of the spallation target with the proton beam variation was investigated. Monte Carlo code FLUKA has been used to calculate neutronic parameters including: spallation neutron flux and energy deposition. As the beam position is shifted towards the positive direction of the Y-axis, the neutron flux and energy deposition increased significantly along the positive direction of the Y-axis, while the flux and energy deposition in the opposite direction decreased. Such a change may lead to a sharp increase of the neutron flux of fuel components near the beam target in subcritical reactor. But as the beam shifts further and further, the neutron flux in the positive direction declines compared to the previous data. The calculation and analyzing the data obtained from proton beam variation shows that it has a great influence on the physical properties of the neutron spallation target.
Periodic Test Period Extension for Partial Closing of Main Steam Isolation Valve
Authors: Liu Deyi , Cao Yong , Zhao Ming , Wang Zilong , Luo Yang , Zou Shengjia , Hu Mengying , Xu Jie , Wang Li 
Published: October 2021

Abstract
The main steam system is an important system connecting the primary and secondary loop of the nuclear power plant. It transports the steam generated by the steam generators to the steam turbine to drive the turbine to turn and do work. The main steam isolation valve (MSIV) is important safety devices in nuclear power plant. Its safety function is to automatically shut down under the accident condition of large break in the main steam pipe or main water supply pipe in order to limit the steam leakage and mitigate the consequences of the accident. The main steam partial shutdown test is a test to periodically verify the operability of the MSIV during the power operation of the nuclear power plant. The test period is usually once a month. However, during the test, there may be shutdown events caused by the mistakenly close of the MSIV, so the test is a high-risk periodic test. At the same time, due to the limitations of field condition and space, there are so many industrial risks and the risk of accidental collision leading to shutdown during the test. At present, the probabilistic safety analysis (PSA) is widely used in operation guidance, maintenance rules and many other fields of nuclear power plants. PSA has gradually developed into an important tool for safety evaluation and decision making. With the development and improvement of PSA technology, risk-informed safety management has been widely used. The this paper, the use of risk-informed PSA method to demonstrate the extension of periodic test period will be introduced. And taking FJS Nuclear Power Plant as an example, the extension of the periodic test period of the partial closing test of the MSIV is demonstrated.
Numerical Study on Multiscale Heat Conduction Problems in Very High Temperature Reactor Fuel Pebble Based on OpenFOAM
Authors: Wang Jincheng , Ding Ming 
Published: October 2021

Abstract
Pebble bed very high temperature reactor (VHTR) has been identified as one of six Generation-IV (Gen-IV) types of reactor which could operate at a high thermal power. The calculation of the temperature in the fuel pebble is a key part of VHTR thermal hydraulics numerical simulation. However, due to the special structure of the VHTR fuel pebble, the temperature calculation involves a multiscale problem. The multiscale heat conduction model includes mesoscale temperature of fuel pebble and microscale temperature of TRISO fuel particles calculation. To deal with the particularity of temperature calculation of the fuel pebble, this paper presents a multiscale heat conduction model based on an open source CFD package OpenFOAM. Firstly, the quasi steady state heat conduction method (QSSHC) and homogeneous layers method (HL) was verified by a simple multiscale model. The results show that the QSSHC method has a good ability of multiscale temperature prediction. Secondly, the mesoscale temperature distribution and the maximum temperature in the microscale of VHTR fuel pebbled are calculated with QSSHC method based on OpenFOAM. This multiscale solver will be couple with other solvers of OpenFOAM, to provide a new perspective of VHTR simulation.
Analysis of IPWR Severe Accident Process Response to SBLOCA
Authors: Yu Hao , Peng Minjun 
Published: October 2021

Abstract
Advanced small modular reactors (SMRs) such as the integrated pressurized water reactor (IPWR) IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. Relap5 / SCDAP/MOD3.4 code is used to model and analyze the severe accident of IP200 reactor caused by SBLOCA superimposed by the emergency power failure. Based on the steady state calculation, which agrees well with designed values, the SBLOCA accident for transient calculation started. First, the SBLOCA accident without the passive core cooling system (PXS) was calculated, and the dynamic process of core degradation in the reactor pressure vessel (RPV) during this process is calculated and analyzed, including transient state response, cooling capacity and thermal hydraulic characteristics. Then the accident situation of different break areas in the small break water loss accident was calculated, and the results were compared and analyzed. The results show that the size of the break area does not affect the sequence of the accident sequence, but only the time and the accident process of the accident sequence. In general, the larger the violation area, the faster the severe accident process. Finally, at the beginning of the exposure in the core, the passive core cooling system (PXS) working can effectively reduce the accident process and protect the integrity of the core and the pressure vessel within the calculation time. Therefore, we can have more time to deal with accidents and effectively prevent the core from melting.
Research and Design of LBB System for Main Pipeline of Nuclear Power Plant
Authors: Jiang Yingying , Xia Hong , Wang Zhichao , Zhang Jiyu , Yin Wenzhe 
Published: October 2021

Abstract
The main pipeline of a nuclear power plant is an important part of the primary coolant pressure boundary. In order to avoid the corresponding nuclear power safety accidents caused by the leakage of the main pipeline, a large number of safety facilities have been built in the design and construction stage of the nuclear power system with the double end shear fracture of the main pipeline as the design basis accident. Complexity and design build costs of nuclear power systems are increased, due to the presence of these safety facilities. It is proved that the possibility of double-end shear fracture of main pipeline is very small by multi-year experience of nuclear power operation. If the cracks and small leakage in the main pipeline of the nuclear power plant can be detected in time and effective measures can be taken, a large number of safety facilities for the double end shear fracture accident can be avoided, the safe and stable operation of the nuclear power plant can be ensured while reducing the complexity and construction cost of the nuclear power system. Based on the above considerations, the LBB system for the main pipeline is studied and designed, which aims to ensure that the cracks and small size leakage caused by the pipeline are monitored in time, so that the occurrence of the cracks and double end shear accidents are avoided.
Investigation of Steam-Injector Operation Mechanism Through Flow Visualization
Authors: Xie X. , Xu Y. , Miwa S. , Sakashita H. , Sawa K. 
Published: October 2021

Abstract
The Steam Injector (SI) is a passive pump with a simple structure that does not require an external power source nor moving machinery for the start-up. The SI constitutes five working parts: the inlet nozzle, mixing nozzle, throat, overflow line, and diffuser. Water and steam are injected into each inlet nozzles, and they flow separately through the mixing nozzle, in which the steam condenses onto the water jet surface. The water jet passes through the throat, and then it flows out from the diffuser where the pressure of the condensed fluid increases. Hence, the SI also possesses the capability as a jet pump, by obtaining a higher discharge pressure than the inlet steam pressure. Furthermore, it has high heat transfer characteristics resulting from direct contact condensation between steam and water. Therefore, SI has a high potential to be utilized as passive safety system that can continuously inject water into the reactor core or auxiliary safety systems in the event of station-black-out. At present, we evaluated the detailed characteristics of SI performances as passive coolant injector from both experimental and analytical approaches.
In this study, axial pressure and temperature measurements with flow visualization of SI were carried out. A detailed database of discharge pressure, internal axial temperature, and pressure distribution, the location of shock wave generation was established by changing the inlet conditions and back-pressure as well as other parameters, to clarify the optimal working conditions. The internal flow behavior of SI was visualized with a high-speed camera, and the images of internal flow characteristics were obtained. Furthermore, the one-dimensional numerical analysis model is improved, and the performance of SI is analyzed according to the experimental results.
Numerical Simulation of Hi Thermal Decomposer in Iodine-Sulfur Cycle Process
Authors: Gao Qunxiang , Wang Laijun , Peng Wei , Zhang Ping , Zhao Gang 
Published: October 2021

Abstract
The very high-temperature gas-cooled reactor is a fourth-generation reactor with inherent safety and modular high-temperature characteristics. Utilizing ultra-high temperature heat to produce hydrogen by iodine-sulfur cycle is one of the important applications. In this process, the decomposition of hydroiodic acid determines the efficiency of hydrogen production, so it is important to design a hydroiodic acid decomposer with the compact structure and good heat preservation. In this paper, numerical simulations are performed on the hydroiodic acid decomposer used for the decomposition of hydroiodic acid. The species transport model, volume reaction and porous model are used to calculate the temperature field and the distribution of each component. The results show that the temperature of the catalytic zone meets the reaction requirements, the temperature drop in the reactor is about 55 K, and the overall decomposition fraction can reach 21.4%. The current flow rate has an acceptable performance for the improvement of the decomposition rate. The results of this study can provide theoretical calculation references for the engineering application of hydroiodic acid decomposers.
Experiment on Vortex Shedding in Water Medium of Three-Way Closed Branch Pipe
Authors: Liu Shuai , Huang Xuan , Feng ZhiPeng , Jiang XiaoZhou , Wang BiHao 
Published: October 2021

Abstract
The vortex shedding phenomenon of the three-way pipe is the main excitation source for pipe vibration, which can lead to the acoustic resonance. In this paper, the vortex shedding phenomenon under the water medium of the three-way closed branch pipe is studied. The test device of a three-way branch pipe with one branch closed is established to verify the mechanism. The PIV measurement method is used to obtain the change characteristics of the flow streamline and velocity distribution of the closed branch pipe under different working conditions. Tests have shown that due to the special structure of the three-way branch pipe with one branch closed, pressure waves are formed at the location of the three-way and vortices are formed in the branch pipe. A large velocity gradient will be generated when the fluid flows through the junction area of the main pipe and the branch pipe. The small vortex clusters that continuously fall off are formed at the front edge of branch pipe that; due to the fluid viscosity, the fluid at the front edge of the branch pipe will be driven downstream by the mainstream area, creating a vacuum area at the front edge. At the interface between the main pipe and the branch pipe, the velocity fluctuation of the vortex increases with the increase of the flow velocity., The size of the vortex cluster in the branch pipe increases with the increase of the velocity and the vortex shedding speed at the front edge of the branch pipe increases with the increase of the velocity.
Study on the Modeling and Simulation of the Horizontal Steam Generator in VVER-1000
Authors: Zhang Ru , Qing Junyan , Bi Xiaolong , Jiang Guanfu , Sun Peiwei , Wei Xinyu 
Published: October 2021

Abstract
Steam Generator (SG) is one of the main components of the power cycle in pressurized water reactor (PWR), and it is the hub of primary coolant circuit and secondary circuit, so the thermal hydraulic analysis of the SG is crucial in the system design and safety analysis of the PWR. The horizontal steam generator (HSG) is one of the main types SG in the PWR nuclear power plant (NPP), and its advantages are that it has more secondary side water capacity and good safety and reliability. The VVER-1000 is a PWR with a thermal power of 3000 MW, and has four HSGs for four loops. The RELAP5 has been used to model the VVER-1000’s HSG and performs the analysis described in this paper. The HSG tube bundle is modeled by three horizontal channels, and the steam control volumes above the heat transfer tube bundle are modeled with three volumes. The steam space is modeled as a steam separator and the steam reception shield is the dryer. The HSG secondary side downcomers are represented with a separate component to provide the power of the natural circulation. To verify the accuracy of the model, three different typical conditions are simulated. The simulation results show that the model built in this paper can correctly simulate the operation of the HSG in VVER-1000.
Study on Radiation Dose Calculation of PWR Spent Fuel Storage and Transportation
Authors: Yang Wen , Li Xing , Qiu Jinrong , Zhou Lun 
Published: October 2021

Abstract
With the rapid development of nuclear energy, spent fuel will accumulate in large quantities. Spent fuel is generally cooled and placed in a storage pool, and then transported to a reprocessing plant at an appropriate time. Because spent fuel is content with a high level of radiation, spent fuel storage and transportation safety play important roles in the nuclear safety. Radiation dose safety are checked and validated using source analysis and Monte Carlo method to establish a radiation dose rate calculation model for PWR spent fuel storage pool and transport container. The calculation results show that the neutron and photon dose rates decrease exponentially with increase of water level under normal condition of storage pool. The attenuation multiples of neutron and photon dose rates are 4.64 and 1.59, respectively. According to radiation dose levels in different water height situations, spent fuel pool under loss of coolant accident can be divides into five workplaces. They are supervision zone, regular zone, intermittent zone, restricted zone and radiation zone. Under normal condition of transport container, the dose rates at the surface of the container and at a distance of 1 m from the surface are 0.1759 mSv/h and 0.0732 mSv/h, respectively. The dose rates decrease with the increasing radius of break accident, and dose rate at the surface of the transport container is 0.278 mSv/h when the break radius is 20 cm. Transport container conforms to the radiation safety standards of International Atomic Energy Agency (IAEA). This study can provide some reference for radiation safety analysis of spent fuel storage and transportation.
Benchmark Study of RBHT Experiment for the Effect of Spacer Grids on Reflood Heat Transfer With LOCUST Code
Authors: Du Qiang , Wen Qinglong , Yuan Hongsheng , Wang Ting , Ruan Shenhui 
Published: October 2021

Abstract
In this study, the system thermal-hydraulic code LOCUST is applied to simulate reflood heat transfer experiments conducted in the RBHT facility, and the effect of spacer grids are considered. The calculation results of LOCUST are compared with experimental data and the calculations of RELAP5 4.0. The results show that both LOCUST and RELAP5 4.0 are capable of predicting the reflood behaviors at a satisfactory level. Besides, the calculations of cladding temperature and heat transfer coefficient are generally in good agreement with experimental data. When spacer grids are introduced, PCT calculated by RELAP5 4.0 and LOCUST are 1178K and 1201K, respectively. However, the calculations without spacer girds are 1184K and 1206K, respectively. The comparison reveals that a decrease of around 5K in the PCT occurs due to spacer grids.
Conceptual Design and Evaluation of Residual Heat Removal System for Small Lead-Bismuth Fast Reactor
Authors: Xu Shijia , Wen Qinglong , Ruan Shenhui , Zhao Ningning , Liu Yukang 
Published: October 2021

Abstract
A high efficient and reliable residual heat removal system (RHRS), which is of great importance in the development of Lead-Bismuth Cooled Fast Reactor (LBFR), was conceptually designed in present study. Based on the design of the RHRS and LBFR, the RELAP5 4.0 code is used to model the system, and then the numerical calculation of steady and transient state was carried out to obtain the important thermal-hydraulic characteristic parameters. Meanwhile, the variations of the parameters were obtained during the transient process, such as the fuel cladding temperature and the natural circulation mass flow rate. The results show that the mass flow rate of the core finally stabilizes at 3.9 kg/s, which is about 1.35% of the rated flow. The peak cladding temperature is less than 750.3 K within 72 h during the whole process, which is far below the temperature safety limit. Therefore, it can be considered that the RHRS can successfully remove the core decay heat of LBFR. This research lays a solid technical foundation for the conceptual design of the RHRS.
Conceptual Design of Nuclear Wet Steam Turbines for Ease of Mass Manufacture
Authors: Ireland Edmund 
Published: October 2021

Abstract
If the nuclear electricity generation is to play a significant role in the decarbonisation of the advanced economies in line with international commitments, construction rates will have to be achieved that are much greater than those achieved historically. This will require a root and branch reassessment of the nuclear technology base to maximise achievable industrial mobilisation. It is important to avoid ‘reactor chauvinism’ that might constrain us to technologies that whilst suited for regular times are un-suited to a challenge of this scale. A conceptual design for a CANDU reactor conventional island is proposed in this work. The design economises on components with laborious fabrication requirements, such as tubed heat exchangers, as well as leveraging developments in other sectors such as power electronics, and its potential performance is evaluated.
A Statistical Approach for Modeling the Effect of Hot Press Conditions on the Mechanical Strength Properties of HTGR Fuel Elements
Authors: Jun Aihara , Masatoshi Kuroda , Yukio Tachibana 
Published: October 2021

Abstract
To maintain the structural integrity of fuel elements for a high temperature gas-cooled reactor (HTGR) under beyond design conditions, strong and oxidation-resistant fuel elements should be further developed. The HTGR fuel elements employ a hot-pressed silicon carbide (SiC)/carbon (C) mixed matrix to improve the oxidative resistance. Hot-press conditions such as pressure, temperature, and duration would be one of the factors that affect the strength of the HTGR fuel elements. To identify the optimal hot-press conditions for preparing the high-strength fuel elements, modeling their effects on the mechanical-strength properties of the HTGR fuel elements should be evaluated quantitatively. In this study, the response–surface model, which represents the relationship between the hot-press conditions and the mechanical-strength properties, has been constructed by introducing statistical design of experiments approaches. The model estimated the optimal hot-press conditions for producing the high-strength fuel elements. We discuss the model’s validity by comparing its predictions with experimental observations.
Study of Recent Sodium Pool Fire Model Improvements for MELCOR Code
Authors: Aoyagi Mitsuhiro , Louie David L. Y. , Uchibori Akihiro , Takata Takashi , Luxat David 
Published: October 2021

Abstract
The Sodium Chemistry (NAC) package in MELCOR has been developed to enhance application to sodium cooled fast reactor. The models in the NAC package have been assessed through benchmark analyses. The F7-1 sodium pool fire experimental analysis is conducted within the framework of the U.S.–Japan collaboration under the Civil Nuclear Energy Research and Development Working Group. This study assesses the capability of the improved models proposed for the sodium pool fire in MELCOR through comparison with the F7-1 experimental data and the results of the SPHINCS code developed in Japan. Pool heat transfer, pool oxide layer, and pool spreading models are improved in this study to mitigate the deviations exhibited in the previous study where the original CONTAIN-LMR models are used: the overestimation of combustion rate and associated temperature during the initial phase of the sodium fire relative to the experimental data and SPHINCS results, and the underestimation of them during the later phase. The analytical result of the improved sodium pool fire model agrees well with the experimental data and SPHINCS results over the entire course of the sodium fire. This study illustrates these enhanced capabilities for MELCOR to reliably simulate sodium pool fire events.
A Beyond Design Basis Earthquake Study of Operating Nuclear Fuel Cycle Facilities
Authors: Liang Li , Rong Pan , Xiuyun Zhu , Guopeng Ren 
Published: October 2021

Abstract
After the Fukushima nuclear accident in Japan, the world attaches great importance to the study of beyond design basis earthquake. In 2016, China National Nuclear Safety Administration issued a new nuclear safety regulation, regulations for design safety of nuclear power plants (HAF102-2016), which requires that appropriate margin must be provided for new nuclear power plants in case of beyond design basis natural disasters. However, the existing nuclear fuel cycle facilities are also faced with the problem of beyond design basis earthquake. However, there is no relevant standard for seismic capacity evaluation of nuclear fuel cycle facilities beyond design basis earthquake in China. In this paper, the seismic safety evaluation standards and evaluation methods of nuclear facilities at home and abroad are investigated. Taking a nuclear fuel reprocessing facility as the research object, the dynamic response numerical analysis under beyond design basis earthquake is carried out. The seismic response is analyzed, and the weak link under the beyond design basis earthquake is found out, and reasonable and feasible seismic margin evaluation is proposed for the nuclear fuel cycle facility. At the same time, it provides technical support for weak links to take reinforcement measures.