bluuebunny/crossref_metadata_2025 huggingface dataset:
A Modified Model for the Net Vapor Generation Point and Its Application on CHF Prediction in Subcooled Flow Boiling
Authors: Akand Md. Abdur Rafiq , Kitahara Kei , Matsumoto Tatsuya , Liu Wei , Morita Koji
Published: October 2021

Critical heat flux (CHF) is one of the most significant thermal criteria for nuclear reactor design and safety in subcooled flow boiling. The accurate prediction capabilities of the characteristic size of departure bubbles are crucial for predicting the net vapor generation point (NVG) and CHF. An experimental research facility was designed to determine the bubble departure diameter and subcooling at the net vapor generation point, not only for vertical flow boiling but also in any orientation between vertical and downward-facing horizontal. An improved force-balanced model is proposed to calculate the bubble departure diameter at the net vapor generation point to predict CHF in vertical subcooled flow boiling. The force-balance equation was solved numerically through iteration to calculate the bubble departure diameter using a novel MATLAB script. The net vapor generation point was modified using the bubble departure diameter obtained from the proposed model. The model agrees well with the experimental data. Finally, the modified departure diameter and NVG were applied to the liquid sublayer dryout model to predict the CHF on upward subcooled flow boiling. It was found that the model predicts the experimental CHF data with an average relative error of 7.17%.
Calculation of Probability of Survival (POS) in Dynamic Systems Based on RMC Code
Authors: Jia Conglong , Shi Guanlin , Feng Zhiyuan , Guo Xiaoyu , Wang Kan , Huang Shanfang , Liang Jingang
Published: October 2021

In a weak neutron source system, the neutron chains rarely overlap. It is necessary to describe this system using stochastic kinetics. Quantities of interest in the field of stochastic kinetics are Probability of Initiation (POI) and Probability of Survival (POS). The RMC code has the capability of calculating the POI in a static system. In this study, the calculation function of POS in dynamic systems is developed in RMC. The Monte Carlo direct simulation method is used for neutron transport in the program. The reactivity of the dynamic system changes with time through the linear interpolation method. The total number of simulated particles was controlled by the threshold and comb method. The benchmark calculated is proposed by the LANL. The calculation results of RMC and MCATK for POI are completely consistent. The calculation results of the two programs for POS are in good agreement except for 600sh(shake). Then the reason for the inconsistent between different calculation results of POS at time 600sh was analyzed and it can be found that the threshold setting in the comb method has a more significant impact on the calculation results of POS. This research is helpful for the further study of stochastic kinetics application.
Dropping-Rod Analysis of Control Rod in ADS Lead-Bismuth Alloy Zero-Power Reactor
Authors: Fu Hui , Lu Daogang , Liu Yu
Published: October 2021

The accelerator-driven subcritical system (ADS) can transmute nuclear waste by using high-energy neutrons, which is the most effective nuclear waste disposal scheme at present. The zero-power lead-bismuth alloy device accurately constructed the interaction mode between nuclear fuel and lead-bismuth alloy coolant material, and more accurately simulated the core physical characteristics of lead-bismuth reactor. Moreover, the inherent safety of the reactor is enhanced by installing several passive safety shutdown systems. However, due to the particularity of the potential harm caused by radioactive material leakage, nuclear-grade equipment should maintain high safety performance under many working conditions, such as earthquake working conditions. To ensure the safe operation of the ADS zero-power device and the safe shutdown under seismic conditions, the motion of the control rod of the ADS zero-power device must be analyzed under seismic conditions. At present, most studies on the control rod drop analysis focus on water reactor, but this paper will analyze and calculate the control rod drop of the lead reactor. In this paper, the physical mechanism of rod falling of the control rod component is analyzed, and the influence of contact collision between the control rod component and the guide tube of the control rod on the whole rod falling time is considered. In this paper, the finite element method is used for transient analysis of ADS lead-bismuth alloy zero power reactor, and the time history diagram of the control bar position is obtained under the seismic conditions, thus verifying that ADS lead-bismuth alloy zero power reactor can be safely shut down under the safe shutdown earthquake (SSE).
Robustness Analysis and Improvement of Fault Diagnosis Model for Nuclear Power Plants Based on Random Forest
Authors: Li Jiangkuan , Lin Meng
Published: October 2021

With the development of artificial intelligence technology, data-driven methods have become the core of fault diagnosis models in nuclear power plants. Despite the advantages of high flexibility and practicability, data-driven methods may be sensitive to the noise in measurement data, which is inevitable in the process of data measurement in nuclear power plants, especially under fault conditions. In this paper, a fault diagnosis model based on Random Forest (RF) is established. Firstly, its diagnostic performance on noiseless data and noisy data set containing 13 operating conditions (one steady state condition and 12 fault conditions) is analyzed, which shows that the model based on RF has poor robustness under noisy data. In order to improve the robustness of the model under noisy data, a method named ‘Train with Noisy Data’ (TWND) is proposed, the results show that TWND method can effectively improve the robustness of the model based on RF under noisy data, and the degree of improvement is related to the noise levels of added noisy data. This paper can provide reference for robustness analysis and robustness improvement of nuclear power plants fault diagnosis models based on other data-driven methods.
RELAP5 Code Analyses of PKL-4 Project Test on PWR Multiple Steam Generator Tube Rupture Accident With Recovery Actions
Authors: Sekine Masashi , Kaneko Junichi , Takeda Takeshi
Published: October 2021

Multiple steam generator tube rupture (MSGTR) accidents are considered in seismic level-1 probabilistic risk assessments (PRAs), because a huge earthquake could cause simultaneous damages to steam generator tubes. Although the occurrence frequency of MSGTR is quite low, the accident may directly lead to a core damage. Therefore, effective recovery actions against MSGTR are essential to avoid the core damage. In order to improve the PRA for MSGTR, the recovery actions should be examined. A MSGTR test with ruptures in multiple loops using the Primärkreisläufe Versuchsanlage (PKL) test facility was conducted in the OECD/NEA-PKL-4 project in order to investigate effectiveness of the recovery actions. The experiment assumed that MSGTR occurred in three out of four loops with a total failure of high-pressure injection system of emergency core cooling system. In the present study, a post-test analysis on the PKL test was carried out with RELAP5/MOD3.3 code (RELAP5). Although discrepancies from the test data appeared in some parameters, the code well predicted the overall trends of the major thermal-hydraulic responses observed in the PKL test. Case studies were also performed to confirm the effects of auxiliary feedwater injection into the secondary side of one or more broken steam generators (SGs) on the core cooling.
The Experimental Research of Surface Characteristics on CHF for the Downward Facing Surface
Authors: Lin Bo , Zhang Lei , Wei Dongshan , Xu Junying , Yun Xiangyu , Zhang Huiyong
Published: October 2021

In-Vessel Retention (IVR) strategy is one of the most important severe accidents mitigation strategies. The Critical Heat Flux (CHF) is an important key element in the assessment of IVR efficiency. How to increase the CHF is a significant method to make sure the IVR success. Heating surface characteristics in an important factor to increase CHF of downward facing boiling surface. This paper aims to study the influence of the different roughness, surface coating, and surface aging on CHF through the mechanism experiments. An experiment facility combining both pool boiling and flow boiling operating modes was designed and taken to carry out series of CHF tests for downward facing surface. It was found that the CHF nearly maintain the same when the roughness Ra achieved a certain value (∼2.0μm), however the CHF with the roughness Ra below this value, the result showed that the nanoparticle deposited surface will enhance the CHF than the smooth surface. The influence of the different surface characteristics had different enhancement, and all surface process enhanced the CHF than smooth surface. It will be helpful to process the surface of the lower head wall when design the IVR management.
Image Based Bubbly Flow Feature Identification Using Deep Learning
Authors: Furuhashi Takashi , Sasaki Takuro , Miwa Shuichiro
Published: October 2021

Gas-liquid two-phase flow has high potential in heat transfer and mixing capabilities, and therefore it is utilized in various technologies such as nuclear reactor and chemical plants. There are several flow regimes since the gas-liquid interface transforms constantly. For the sake of safety and optimization in operating plants, it is crucial to understand the behavior of the gas-liquid interface. We have focused on extracting the bubble features in the bubbly flow by filming the bubbly flow with a high-speed camera and training convolutional neural network (CNN) for feature extraction. The assumption made was bubbles in the bubbly flow being ellipsoids. Since void fraction and interfacial area concentration are one of the geometric parameters in the two-phase flow models, like two-fluid model, it becomes possible to evaluate the flow field of the two-phase flow quickly and quantitively by calculating these parameters from the extracted features. We have compared two-phase flow parameters with the conventional object detection method using bounding boxes, and the new ellipse fitting method to identify the best region proposal shape. As a result, the conventional method showed higher accuracy in extracting bubble features under our flow conditions.
The Interfacial Area Weighted Area-Averaged Gas Velocity Model for the Interfacial Area Transport Equation in the System Analysis Code
Authors: Shen Mengsi , Lin Meng
Published: October 2021

The interfacial area transport equation is a more accurate and stable way to compute the interfacial area concentration than the traditional empirical correlation in the two-phase two-fluid model. And among the parameters in the two-group interfacial area transport equation, the interfacial area concentration weighted area-averaged gas velocity is an important parameter to close the two-group area-averaged interfacial area transport equation in the system analysis code. However, there has been no theory model to compute the interfacial area concentration weighted area-averaged gas velocity until now. So this study established the theory model for two-group interfacial area concentration weighted area-averaged gas velocity based on the drift-flux model for the two-phase dispersed bubble flow. The experimental data were selected from the published literature, which include the detailed two-phase interfacial structure experimental data for the slug bubble flow. The interfacial area concentration weighted area-averaged gas velocity model predicted the selected experimental data well, which validated the developed model. Moreover, the difference between the interfacial area concentration weighted area-averaged gas velocity and the void weighted area-averaged gas velocity is clarified quantitatively for the first time. The theory model developed in this study can be improved and then be used to compute the interfacial area weighted area-averaged gas velocity because it includes the empirical parameter of conventional drift-flux model.
Key Parameters Determination of Integral-Plate Cruciform Control Rod
Authors: Zhang Hao , Li Songyang , Wang Dingqu , Jiang Yueyuan , Hao Wentao , Xiong Wei , Ma Jizhong
Published: October 2021

Integral-Plate cruciform control rod is a new type of control rod designed and manufactured by the Institute of Nuclear and New Energy Technology (INET), Tsinghua University. Cruciform control rod is assembled by two longitudinal slotted absorber plates which are perpendicular and inserted into each other. Hundreds of holes are drilled, symmetrically from each flank of the wing plates to the center. Most holes are designed to contain B4C absorber elements, and the other ones are used to connect the gas space. Compared to the conventional “absorber-rod & cladding-sheet” design, the new type control rod has more B4C loading amount under the same height of absorption zone. The Integral-Plate design can effectively reduce the overall thickness and weight, improve the mechanical performance, and cut down the cost and manufacturing difficulty. Plate thickness, hole diameter, hole spacing and other parameters determine the neutron absorption capacity and structural characteristics of the control rod, which are the key parameters affecting the performance and economy of the Integral-Plate cruciform control rod. In the design process, constraints of the engineering technology should be considered, and the economy should be improved as much as possible on the premise of ensuring the performance. In this paper, a method of determining the parameters such as diameter and spacing of absorber holes under a series of different plate thickness is established by means of performance and mechanical analysis, which provides convenience for the design of Integral-Plate cruciform control rod in different reactor core parameters.
Feasibility Study of Disassembly Technologies of Fast Reactor Fuel Assembly
Authors: Nishikawa Hidetsugu , Takeuchi Masayuki , Kitagaki Toru , Tooya Yuuichi
Published: October 2021

In the development of disassembly technologies for the fast reactor fuel assembly, laser cutting was used in tests done in the United Kingdom and the United States due to problems of wear and breakage of cutting-off wheels (CWs); however, the laser can cause thermal damage to fuel pins, melting the part of the fuel pins under the wrapper tube and releasing radioactive materials.

Laser cutting conditions to mitigate thermal damage of fuel pins, and development of CWs with appropriate durability to withstand wear and breakage damage were studied. A mock-up machine to be used with a simulated fuel assembly of the fast reactor Monju was designed and manufactured, and the test to confirm the proposed disassembly procedure was conducted. Laser cutting with the reviewed cutting conditions, mechanical cutting with controlled feed speed using CWs, and the disassembly procedure tested with the mock-up machine were confirmed to be workable disassembly technologies.
The Safety of Nuclear Fuel Cycle Facilities in China After the Fukushima Accident
Authors: Que Ji , Yang Xiao-wei , Liu Yun-tao , Zhao Shan-gui , Shen Hong , Liu Tian-Shu
Published: October 2021

The Fukushima Accident happened 10 years ago. During the 10 years, the nuclear safety regulatory body of china takes some activities in non-reactor nuclear fuel cycle facilities, responding to the impact of the Fukushima accident. The activities include those review activities initiated immediately after the accident and those regulations revision activities. They will be presented in the present paper. The approaches for review, so called the safety reassessment, were firstly worked out. Those included a walk down with experts through the facility, the update of the external natural events, the review or revalidation of the design basis for the external natural events, the emergency preparedness and response for extreme natural disasters. According to the approaches, operating organizations implemented the safety reassessment to facilities which they operated and formed the result reports. The review results of those reports should be summarized into two points. The first point was that the information of the external natural events for each site was up-to-date, because new facilities continue to be built in the same site. The second point was the some building structures did not meet the requirements, because some standards were out of date, or updated. Meanwhile, the regulatory body thought that it was necessary to make a graded approach for different facilities. Finally, a rule was made. the graded approach also will be present in the paper.
The Radioactivity Monitoring of Environmental Samples in Zhejiang During the Events of Nuclear Leakage in Japan
Authors: Liu Gongye , Yang Jia , Hu Xiaoyan , Hu Fei , Xiang Yuanyi
Published: October 2021

This article introduces the emergency monitoring of environmental samples during the events of nuclear leakage in Japan in March 2011. 17 environmental samples were collected, and the concentration of γ emitting radionuclides was analyzed. The monitoring results show that small amounts of artificial radionuclides, 131I and 137Cs, were detected in pakchois and spinach on April 7. We also compared the measurement results with average values of the radionuclide activity concentrations in the same samples over Zhejiang province in the recent ten years.
Review of the Configuration Risk Management Methodologies
Authors: Zhang Yuhang , Zhang Zhijian , Wang He , Zhang Lixuan , Sun Dabin
Published: October 2021

To ensure nuclear safety and prevent or mitigate the consequences of accidents, many safety systems have been set up in nuclear power plants to limit the consequences of accidents. Even though technical specifications based on deterministic safety analysis are applied to avoid serious accidents, they are too poor to handle multi-device managements compared with configuration risk management which computes risks in nuclear power plants based on probabilistic safety assessment according to on-going configurations. In general, there are two methodologies employed in configuration risk management: living probabilistic safety assessment (LPSA) and risk monitor (RM). And average reliability databases during a time of interest are employed in living probabilistic safety assessment, which may be naturally applied to make long-term or regular management projects. While transient risk databases are involved in risk monitor to measure transient risks in nuclear power plants, which may be more appropriate to monitor the real-time risks in nuclear power plants and provide scientific real-time suggestions to operators compared with living probabilistic safety assessment. And this paper concentrates on the applications and developments of living probabilistic safety assessment and risk monitor which are the mainly foundation of the configuration risk management to manage nuclear power plants within safe threshold and avoid serious accidents.
Assessment of Seismic Fragility Using a Three-Dimensional Structural Model of a Reactor Building
Authors: Nishida Akemi , Choi Byunghyun , Shiomi Tadahiko , Kawata Manabu , Li Yinsheng
Published: October 2021

New regulatory requirements in Japan have strengthened the mitigation of damage caused by natural hazards, such as earthquakes, and the operational guide for safety improvement evaluation recommends the use of a probabilistic risk assessment (PRA) as the evaluation method in Japan. In the PRA of an earthquake, also known as the seismic PRA, one of the most important issues is the realistic assessment of the structural seismic response and the conditional damage probability (fragility) assessment using a realistic response assessment of nuclear buildings and equipment. Accordingly, we conducted this study both on the analysis methods used for realistic seismic response and assessment methods of seismic fragility to ensure the seismic safety of nuclear buildings and equipment. In this study, we use a three-dimensional (3D) structural model of a reactor building to conduct a nonlinear seismic response analysis for input ground motions beyond a design basis. In addition, we identify the failure mode of the structural components of the reactor building associated with the equipment and assess the seismic fragility based on the 3D behavior of the reactor building. The local response and detailed damage progress of the reactor building obtained through seismic response analysis are reported, along with the results of the seismic fragility assessment.
Level 2 Probability Safety Assessment for External Events: Approach and Application for NPPs in China
Authors: Liu Yu , Wang Cong , Liu Jing , Gao Heng
Published: October 2021

Fukushima accident is a tragic case that extreme external events (tsunami and seismic) threaten the safety of NPP and lead to large uncontrollable release of radioactive materials. Therefore, how to perform the study on external events Level 2 PSA has become a crucial problem needed to be considered deeply.

In this study, the main purpose is to describe an external events Level 2 PSA approach and apply it for a III generation PWR in China. This study consists of the following factors:

(1) Screening out external events.

(2) Focusing on the key technical elements of Level 2 PSA with special considerations on the characteristics of different external events.

(3) Building a detailed Level 2 PSA model for the screened events with software RiskSpectrum and obtaining the quantitative results of large release frequency (LRF).

(4) According to the analysis results, finding out the risk weaknesses and summarizing recommended proposals.
Summary of the Practice of Clearance of Uranium-Containing Calcium Fluoride Slags in China’s Nuclear Facilities
Authors: Jiang Jing , Lei Qiang , Xu Chen , Zhu Zhaowen , Xu Chunyan , Wang Shijun , Li Xiaolong , Zhang Min
Published: October 2021

During the operation of some nuclear facilities in China, a large amount of uranium-containing and fluorine-containing wastewater will be produced, which requires de-uranium and de-fluorine treatment before discharge. The commonly used method for removing fluoride at present is the chemical lime precipitation method, that is, adding lime to fluorine-containing wastewater to combine fluoride ions with calcium ions to form calcium fluoride precipitation that is insoluble in water. This method has the advantages of low cost and easy operation, but it will produce a large amount of uranium-containing calcium fluoride slag. The content of CaF2 in these slags is more than 80%. If the uranium-containing calcium fluoride slag is cleaned and controlled and reused as a non-radioactive waste, it will not only use fluorine resources, but also eliminate the secondary pollution of fluorine.

This article mainly introduces the clearance activities of uranium-containing calcium fluoride slags carried out by the nuclear facilities, including process conditions, source item investigation, waste sampling, nuclide concentration analysis, release control evaluation, release destination and quality Guarantee, etc., and provide relevant management and technical advice on the cleaning and control of uranium-containing calcium fluoride slag based on the above-mentioned practice.
Security Analysis Based on Probabilistic Safety Analysis Coupled With Deterministic Safety Analysis Used RAVEN
Authors: Sun Dabin , Zhang Zhijian , Li Lei , Wang He , Chen Sijuan , Zhang Yuhang , Zhang Lixuan
Published: October 2021

Probabilistic safety analysis methods (PSA) and deterministic safety analysis methods (DSA) are classical safety analysis methods for nuclear power plant (NPP). However they can’t do a comprehensive analysis for epistemic uncertainty and aleatory uncertainty. This problem can be solved by Risk-Informed Safety Margin Characterization method (RISMC) which combines PSA and DSA.

Due to Best-Estimate plus uncertainty analysis (BEPU) analysis method quantify the epistemic uncertainty in safety analysis accurately, BEPU is used for DSA. The dynamic event tree (DET) analysis method can branch the event tree according to the number of successfully invested devices, which can provide a more detailed component accident sequence, which is of great significance for serious aleatory uncertainty. In order to improve the analysis efficiency, the RAVEN platform is used for safety analysis.

The uncertainty of NPP safety analysis is analyzed more accurately by using the RISMC analysis method which combines PSA and DSA. The CDF calculated by the BEPU analysis method coupled with the DET method is 3.59e−12, and the CDF obtained by the DET analysis method is 2.59E−15. Moreover, the efficiency of the analysis is greatly improved by using the RAVEN platform and the method of selecting significant sequence analysis.
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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.