bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Experimental Research on the Mechanical Characteristics of the Newly Designed Fuel Rod Supporting Structure of Spacer Grid
Authors: Ren Quan-yao , Pu Zeng-ping , Zheng Mei-yin , Chen Jie , Qin Mian , Han Yuan-Ji , Peng Yuan , Chen Ping , Zhao Rui-rui , Li Jian-wei 
Published: October 2020

Abstract
The fuel rod supporting structure of spacer grid is the key to ensure the horizontal and axial position of the fuel rod. On the purpose of improving the neutron economy and simplifying the manufacturing process of the spacer grid, the innovative curved dimple and arched spring have been designed, which could be directly punched from the strip of spacer grid. The mechanical experiments have been performed to acquire the deformation-load curves of dimples and springs on single strap and in-cell supporting structure, which could provide the load and residual deformation at 100% and 120% designed nominal deformation. It has been demonstrated that the designed in-grid cell has a relative stable load at the 100% nominal designed deformation and the test method of the single stripe is representative for the similar load-deflection curve between the spring and the in-grid cell.
Research on Fault Diagnosis of Reactor Coolant Accident in Nuclear Power Plant Based on Radial Basis Function and Fuzzy Neural Network
Authors: Sun Pengpeng , Liu Yong , Wu Guohua , Duan Zhiyong 
Published: October 2020

Abstract
Nuclear power plants (NPPs) are widely used in the world. After three nuclear accidents, people propose higher of the safety and reliability on NPPs. Reactor coolant system (RCS) in the NPP directly affects whether the heat can be exported and radioactivity can be inclusive. It plays an important role of the NPPs safety. So, it is great significance of fault diagnosis for RCS in NPP.
Although many scholar had carried out research on fault diagnosis of NPPs, different networks may lead to different results in a system. Therefore, this paper chooses a system and uses different neural networks (NN) for comparative analysis which can provide advice for follow-up research. In the paper, RCS has been analyzed and typical fault have been analyzed through PCTRAN simulator. On this basis, two kinds of NN combined with fuzzy systems: radial basis function (RBF) and back propagation (BP) are used for fault diagnosis and comparative analysis. Loss of coolant accident, single pump failure, loss of feed water are set for simulation experiment. Simulation experiment shows that BP network’s hidden layer nodes is less than RBF-NN, but iteration speed of BP network is faster; accuracy of fault diagnosis based on BP-NN is higher than RBF-NN; fuzzy-NN for fault diagnosis is faster than NN.
Implementation and Validation of an Aerosol Collection Model by a Spray in a CFD Code: Application to the Scavenging of Aerosols Released During Laser Cutting Operations of Fuel Debris for the Dismantling of the Damaged Reactors of Fukushima Dai-ichi
Authors: Gelain Thomas , Porcheron Emmanuel , Leblois Yohan , Doyen Ioana , Chagnot Christophe , Journeau Christophe , Roulet Damien 
Published: October 2020

Abstract
The general context of this article is related to the dismantling of the damaged reactors of Fukushima Dai-ichi and, more specifically, to the implementation of the laser cutting technique for the fuel debris retrieval. IRSN is involved in a project led by ONET Technologies and in partnership with CEA, to bring relevant elements in order to analyze the risks induced by the dispersion of aerosols released by the dismantling operations.
During the laser cutting operations in air or underwater conditions, particles will be produced, involving a potential risk of dispersion into the environment. Hence, in order to prevent this situation, their collection is one of the safety key issues in the in-situ dismantling actions. For that, IRSN performed CFD simulations of aerosol scavenging by a spray to evaluate the collection efficiency by this technique.
These simulations, conducted with the ANSYS CFX code, use an Eulerian method for the continuous phase, and a Lagrangian method for the spray for which a collection model detailed by Plumecocq [1] or Marchand [2] was implemented. Aerosols are modelled by a DQMOM population balance implemented by Gelain et al. [3] (already used for recent simulations in the same context), and enriched with a deposition model developed by Nerisson et al. [4].
At first, CFD simulations were performed with the geometry of the IRSN TOSQAN facility [5], comparatively to experimental results presented in a previous paper [6]. This step enables the validation of the collection model implementation and to study the sensitivity to the aerosol size.
Then, CFD simulations were conducted with the geometry of the pedestal of Fukushima Dai-ichi reactors, to be more representative of a realistic case. For this configuration, sensitivity studies are described, highlighting both the influence of a multispray and of thermal-hydraulic conditions (temperature) on aerosol scavenging efficiency.
Study on the Spacer Grid Dynamic Crush Strength According to Cell Sizes
Authors: Ryu Jooyoung , Woo Hangil , Nam Chaeyoung , Yoo Youngik , Park Namgyu , Lee Seungjae 
Published: October 2020

Abstract
A spacer grid is one of the primary components of the PWR nuclear fuel. Spacer grid maintains proper pitches between the fuel rods and enables the fuel rod to cool down by providing coolant flow path. However, when the nuclear fuel is subjected to an unexpected excessive load during shipping, handling, manufacturing and operating, it could lead to fuel failure such as spacer grid buckling and cladding tube deformation. The most limiting load acting on the spacer grid is the lateral impact load during seismic/loss-of-coolant accidents. Dynamic crush strength of the spacer grid greatly contributes to the nuclear fuel integrity throughout the fuel lifetime [1]. This buckling strength tends to become weak in end of life (EOL) condition. KEPCO NF (KNF) carried out dynamic crush tests of the spacer grid and analyzed its characteristics. Spacer grids were prepared with three groups that have different cell sizes according to beginning of life (BOL), EOL and enlarged EOL simulated conditions. In addition, two kinds of dynamic crush tests were performed. One is pendulum impact test that drops a hammer to the grid in a short time. And the other is hydraulic long-pulse test that pushes impact plate to the grid in longer time. These tests and analysis results were compared in each group and discussed to explore key factors for improving crush strength of the spacer grid. In this paper, the spacer grid manufactured by additive manufacturing (AM) technology [2] is also introduced to verify the buckling performance. AM is a method to make designed shape with metal powder and built-up technology that is different from conventional manufacturing. Through the study, it could be a good alterative solution that the new manufacturing method might be helpful to improve dynamic impact characteristics.
Study of Laser 3D Scanning Model Reconstruction for Nuclear Facilities Decommissioning
Authors: Yongling Zhang , Wan Zhao , Shuai Wang 
Published: October 2020

Abstract
The laser 3D scanning model reconstruction is an effective technical means for the reconstruction of 3D virtual scenes of old nuclear facilities. However, the existing technology have the following problems: First, the scanning equipment needs to be installed on the triangular fixed bracket, which requires personnel to be installed on site, thereby bringing radiation safety risks to personnel; second, it requires the person to reconstruct and splice the model one face by one face, which is inefficient. Aiming to deal with these problems, this paper proposes a mobile laser 3D scanning method and a rapid model reconstruction technology. These methods include: (1) In order to realize fast scanning of large-scale scenes, and at the same time ensure accurate scanning of local important objects, a laser three-dimensional composite scanning method is proposed. (2) Design and development of a remote control automatic walking and fixing device, which can transport the laser 3D scanner to the position of measurement which is needed. The device has a certain ability to overcome obstacles, and can be reliably fixed to ensure the stability and scanning accuracy of the laser scanner at high speed working station. You can remotely control the laser scanner to start or stop working. The difficulty of this technology is how to make sure the device has the flexibility in motion and the stability on working; (3) An automatic model reconstruction and feature identification method based on a large number of point cloud data is proposed, and the corresponding software is developed. This method can realize the automatic recognition of the standard and common models, and the automatic model reconstruction of the point cloud data in the scene. To achieve this goal, the neural network algorithm is used. These technologies can effectively reduce the radiation safety risk of personnel during laser scanning in high-radioactive places, reduce the intensity of personnel operations, and improve work efficiency.
Point Defect Effects on Tensile Strength of BCC-Fe Studied by Molecular Dynamics
Authors: Lin Pandong , Nie Junfeng , Liu Meidan 
Published: October 2020

Abstract
BCC-Fe is the critical and major component of the reactor pressure vessel (RPV) steel. With long-tern neutron irradiation, many point defects can be obtained in RPV steel. In this paper, the points defects (interstitial, vacancy and Frenkel pair) effects on the tensile strength of Fe are studied by molecular dynamics simulations at 300K. The uni-axial tensile load is along [001] direction of the Fe samples loading in constant strain rate. The Fe atoms are added or removed randomly to generate point defects. For point defects, three types of point defects can decrease the tensile strength containing yield stress and strain of Fe samples. In addition, the tensile strength decreases with the increase of point defect concentration. With the same defect concentration, interstitials decrease the yield stress the most seriously compared with the vacancies and Frenkel pairs. Apart from that, the morphology and evolution of the microstructure of Fe with point defects are also investigated under tension. Compared with the perfect crystal, the generation of dislocation decreases the tensile strength dramatically. For sample with interstitials, interstitial clusters form and evolve in dislocations loops finally. For sample with vacancis, vacancy may aggregate together and vacancy clusters form as a result, which is seen as precursors of dislocation loop. Notably, the results are meaningful to understand the effects of point defects on tensile strength of BCC-Fe.
Effects of Rolling on Characteristics of System Under Forced Circulation and Natural Circulation
Authors: Cai Rong , Yue Nina , Fang Hongyu , Chen Baowen , Liu Lili , Ma Zehua 
Published: October 2020

Abstract
The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.
Nuclear Facility Safety at the United States Department of Energy
Authors: Frias Patrick , Muñoz José R. O. , Restrepo Louis , Tingey James L. , Louie David L. Y. 
Published: October 2020

Abstract
Nuclear facility safety is crucial to preventing and/or reducing high consequence-low probability accidents and, thus reducing the potential risks posed by United States Department of Energy (DOE) and National Nuclear Security Administration (NNSA) operations at their facilities/activities. DOE/NNSA has the responsibility of developing, issuing, maintaining, and enforcing nuclear safety Directives while fostering a culture that promotes nuclear safety research and development. Lessons learned from past accidents, near misses, and experiments/analyses are also important resources for improving operational nuclear safety in the safety community.
This paper first identifies and describes the current Directives in place, including safety review and regulatory process, and safety programs that support implementation of the Directives. This paper also describes a contractor’s approach to identifying and implementing safety using these Directives and lessons-learned in multiple discipline areas of nuclear safety.
Flow Pattern Transition Criteria for Upward Two-Phase Flow in Annulus
Authors: Wen He , Chenru Zhao , Hanliang Bo 
Published: October 2020

Abstract
Vertical upward two-phase flows in annulus are of great importance in many industrial fields due to the closely relationship between the flow patterns and the heat transfer characteristics. Common flow patterns in annulus are bubbly (B), slug (S), churn (C) and annular (A) flow, most of which are quite similar to those in tubes. However, due to the elliptic nose and asymmetric shape of the Taylor bubble in annulus, the slug to churn flow transition could be influenced by the channel geometry which was usually ignored in most of the previous researches. The flow pattern transition criteria for tubes are thus not applicable for annulus, especially for slug to churn flow transition, which should be separately studied. Therefore, in this paper, the basic characteristics of the flow pattern in annulus and their transition mechanism are analyzed. In addition, a set of semi-empirical transition criteria with higher accuracy are assessed and selected for annulus based on theoretical analysis and comparisons with experimental data.
Research on Fault Diagnosis Based on Hierarchical Signed Directed Graph for Nuclear Power Plants
Authors: Guohua Wu , Jiaxin Wang , Diping Yuan , Yiqing Xiao 
Published: October 2020

Abstract
When nuclear power plants (NPPs) are in failure state, it may release radioactive substance into the environment. Thus, the safety of NPPs is put forward a high standard. Fault detection and diagnosis (FDD) are significant for NPPs to help operator timely know the state of system and provide the online guidance. Fault diagnosis can improve the safety of nuclear power plants, but current fault diagnosis methods pay too much attention to accuracy of diagnostic results. As a complex industrial system, how to explain the causes of faults in NPPs becomes more important. Although there are many studies on the knowledge graph, there is no detailed description on the failure process (consider timing). This paper proposed a three-layer structure for FDD in NPPs. Each layer represents the stage of the accident, it can give the operator a clear cognitive process to faults. The three-layer structure includes “smooth layer”, “threshold layer” and “fault layer”. The three layers indicate the reason of faults, the response of the parameters at each stage, and clearly showing the accident process. The smooth layer uses the stability analysis to analyze whether the current NPP is operating abnormally; the threshold layer uses the thresholds of the NPP to monitor which parameters have exceeded the upper limit or the lower limit; the fault layer reflects what is happening in the current operation and accidents are explained using signed directed graph. This paper takes the loss of coolant accident as an example, three-layer structure is analyzed, which shows the feasibility of the method. The case shows that the proposed method is superior to the conventional SDG method, can diagnose the faults, and give the reason of diagnosis results.
Methodology Development for Transient Flow Distribution Analysis in High Temperature Gas-Cooled Reactor
Authors: Aoki Takeshi , Sato Hiroyuki , Ohashi Hirofumi 
Published: October 2020

Abstract
In the thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR), unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed to minimizing the unintended flows. The flow distribution considering unintended flows in the reactor has been evaluated for steady and conservative condition. On the other hand, the transient thermal hydraulic analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR.
The present study aims to improve the thermal hydraulic system analysis code developed by Japan Atomic Energy Agency based on the RELAP5/MOD3 code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents for its design improvement utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated to evaluate the detailed flowrate distribution considering the unintended flows in the core and the molecular diffusion that is important to analyze beginning air ingress behavior in an air ingress accident triggered by a rupture of a primary coolant piping in HTGR. It is concluded that a prospect has confirmed to apply the improved thermal hydraulic system analysis code for transient flow distribution analysis for prismatic-type HTGRs.
Experimental Study of Pure Steam and Steam-Air Mixture Condensation on a Vertical Chrome-Plated Tube
Authors: Niu Zesheng , Fan Guangming , Cheng Jie , Li Wei 
Published: October 2020

Abstract
To investigate the heat transfer characteristics of the chrome-plated tube and hope it could be used on the internal heat exchanger of passive containment cooling system (PCS), an experimental investigation has been conducted. In this experiment, a series of steam condensation experiments are performed under pure steam and steam-air mixed conditions over chrome-plated tube for a variety of chromium coating thickness (1μm and 10μm), total pressure, air mass fraction and wall subcooling. Condensation heat transfer coefficient was obtained for the total pressure ranging from 0.2 MPa to 0.4 MPa, air mass fraction ranging from 0.10 to 0.71, and wall subcooling from 10°C to 70°C. Moreover, the designed visualization experimental device makes the experimental phenomenon can be directly observed through the observation window. Under the pure steam condition, the results show that droplet condensation and filmwise condensation is co-existed on both two kinds of chrome-plated tubes, the chrome coating thickness of 10μm tube shows better heat transfer ability. Under the steam-air mixed condition, the condensation heat transfer coefficient of both two kinds of tubes increases with total pressure, and decrease with the air mass fraction and wall subcooling, while the influence of chrome coating thickness on heat transfer is no longer noticeable. The results also indicate that the thickness of the chromium coating will affect the surface microstructure of the chrome-plated tube and then affect the heat transfer ability of the chrome-plated tube.
The Fire Event Analysis for Fire Frequency Estimation on Japanese Nuclear Power Plant
Authors: Nagata Yasunori , Uchida Tsuyoshi , Shirai Koji 
Published: October 2020

Abstract
To promote the practical fire Probability Risk Assessment (PRA) among Japanese industries based on the domestic fire experience, it is important to estimate generic fire ignition frequency distributions for Japanese nuclear power plants (NPPs), which are applicable to the fire PRA during full power operation.
Firstly, the Central Research Institute of Electric Power Industry (CRIEPI) proposed the estimation process of the distributions, which consists of the four steps including sub steps as follows:
Step 1: Fire event data collection including development of criteria
Step 2: Fire event data analyses including development of criteria:
Step 2.1: Screening of fire events to identify events to be classified
Step 2.2: Classification of fire events from the viewpoint of fire severity of each event.
Step 2.3: Categorization of fire events from the viewpoint of fire ignition source types
Step 3: Estimation of generic fire ignition frequency distributions
Step 4: Comparison of the results of Step 3 with fire ignition frequency estimated in other countries
In this paper, CRIEPI introduced the detail of Step 1 and 2 above and discussed the study results for them, which are essential to estimate generic fire ignition frequency distributions properly. Through Step 1, CRIEPI established a preliminary fire event database based on fire records at Japanese NPPs provided by domestic utilities. In step 2.1, CRIEPI identified the 180 fire events for future steps. Finally, CRIEPI classified the fire events and defined the fire ignition source categories in Step 2.2 and 2.3. As a result, CRIEPI found the major three risk contributors among fire ignition source categories: transient related fires, electrical cabinet fires and pump fires.
Thermo-Mechanical Analysis of Instrumentation Guide Tube Failure During a Severe Accident in a Nordic Boiling Water Reactor
Authors: Yue Ying , Villanueva Walter , Wang Hongdi , Wang Dingqu 
Published: October 2020

Abstract
Vessel penetrations are important features of both pressurized water reactors and boiling water reactors. The thermal and structural behaviour of instrumentation guide tubes (IGTs) and control rod guide tubes (CRGTs) during a severe accident is vital in the assessment of the structure integrity of the reactor pressure vessel. Penetrations may fail due to welding failure, nozzle rupture, melt-through, etc. It is thus important to assess the failure mechanisms of penetrations with sufficient details. The objective of this paper is to assess the timing and failure modes of IGTs at the lower head during a severe accident in a Nordic boiling water reactor. In this study, a three-dimensional local finite element model was established using Ansys Mechanical that includes the vessel wall, the nozzle, and the weld joint. The thermo-mechanical loads of the finite element model were based on MELCOR results of a station blackout accident (SBO) combined with a large-break loss-of-coolant accident (LBLOCA) including an external vessel cooling by water as a severe accident management strategy. Given the temperature, creep strain, elastic strain, plastic strain, stress and displacement from the ANSYS simulations, the results showed the timing and failure modes of IGTs. Failure of the IGT penetration by nozzle creep is found to be the dominant failure mode of the vessel. However, it was also found that the IGT is clamped by the flow limiter before the nozzle creep, which means that IGT ejection is unlikely.