bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.