bluuebunny/crossref_metadata_2025_split huggingface dataset:
Code to Code Validation of SAC-3D Based on EBR-II Benchmark Problem
Authors: Lu Daogang , Lyu Siyu , Sui Danting 
Published: October 2020

Abstract
A lot of previous works indicated that the 3D neutronics calculation module can provide better prediction results in reactor system safety analysis especially under several anticipated transient accidents which with strong spatial influence factors. SAC-3D code is a new version of SAC series reactor system safety analysis codes which was developed by North China Electric Power University[4][5][6][7]. In this version, the neutronics calculation module was developed based on the three-dimensional nodal expansion method. In 2012, IAEA initialed a CRP on benchmark analysis of EBR-II loss of flow without scram test, SHRT-45R benchmark data is one of the benchmark specifications provided by ANL. In the present work, the neutronics benchmark problem of EBR-II SHRT-45R was analyzed with SAC-3D. The neutronics calculation module of SAC-3D was validated by comparing the key results with the benchmark data. The simulation results also agreed well with the results provided by other participants of this CRP.
Stress Analysis of the Lower Head of Central Measuring Shroud Under Thermal Striping and Thermal Shock Conditions
Authors: Zheng Shu , Lu Daogang , Cao Qiong 
Published: October 2020

Abstract
The central measuring shroud, as an important in-vessel component, provides guidance and protection for control rods and measuring equipment in a sodium-cooled fast reactor (SFR). The lower head of central measuring shroud (LHCMS), which is located above the core outlet, is only 500mm away from the core outlet. Therefore, the LHCMS is affected by the liquid sodium from core outlet for a long period, especially the temperature effects of the following two types. On the one hand, under the operating condition of the SFRs, the uneven distribution of the core power causes the phenomenon of thermal striping, which may cause high cycle fatigue and even initial crack. On the other hand, under the scram condition, the coolant temperature at the core outlet is sharply reduced due to the decrease of the core power, inducing the phenomenon of thermal shock that may cause large thermal stress and low cycle fatigue. Therefore, stress and fatigue analyses of the LHCMS under the thermal striping and thermal shock conditions are very necessary. In the paper, finite element model of the LHCMS was first established, and then according to the temperature curves under thermal striping conditions and thermal shock conditions, the thermal stress of the LHCMS was simulated. The results showed that although the temperature fluctuation outside the LHCMS is severe, the stress caused by thermal striping only slightly fluctuates at 123MPa level, the maximum stress range is 11MPa. Besides, at 20s, there exists the maximum stress difference between thermal striping and thermal shock conditions, the maximum stress caused by thermal shock is about 3 time larger than that caused by thermal striping. According to high cycle and low cycle fatigue analyses, the fatigue damage factor of thermal striping is only 0.0078, while the fatigue damage factor of thermal shock is 3.416, which should provide a reference for the design of the LHCMS.
Synergy of an SMR for Addressing Remote Communities Non-Nuclear Waste
Authors: Harvel Glenn 
Published: October 2020

Abstract
SMRs are a popular topic with a significant number of designs with a wide range of sizes. There has been some assessment done with respect to non-electrical applications including district heating and desalination. The drive of this type of work is to find alternative uses for the SMRs so that the thermal energy is more effective and hence the SMR is more economical. This work is similar in that it studies the synergy that might exist between a remote community and the SMR. Most work for SMRs related to remote communities with the impact as one-way, that is the benefit of the SMR to the community yet the SMR is a separate plant. The consideration here is that the SMR could be used to burn the non-nuclear waste products of the community and return useful products. An example is the plastics generated by the community can be converted into a usable fossil fuel, such as kerosene, by using the heat energy of the SMR. The SMR then has a dependency on the community waste stream. In this manner, the environmental load of the community is reduced yet the community also obtains a locally produced fuel that could be used for heating or transport outside of the community. Considering that diesel fuel costs can be extremely high in remote communities, methods to reduce the fuel costs, including manufacture of their own fuel, can result in a synergistic or symbiotic relationship between the community and the SMR and the community can then have a centralized energy area for supporting neighbouring communities.
An Integrated Experimental Test Facility to Support Development of the Passive Containment Cooling System of HPR1000
Authors: Li Wei , Du Shuhong , Gu Weiquan , Zhang Nan , Ding Ming , Sun Zhongning , Meng Zhaoming 
Published: October 2020

Abstract
HPR1000 is an advanced nuclear power plant with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. It is based on the large accumulated knowledge from the design, construction as well as operations experience of nuclear power plants in China. The passive containment cooling system (PCS) of HPR1000 is an important and innovative passive safety system to suppress the pressure in the containment during LOCA. In this paper, the detailed design process of PCS is reviewed, and an integrated experiment facility for the study on the coupling behavior between PCS and thermal hydraulic characteristics in the containment is described, and arrangement of measuring points including temperature, pressure, gas composition and so on are introduced in detailed. Also, the experimental energy released and energy vent to ensure the similarity of containment pressure response, thermal stratification and PCS heat removal is introduced. According to this versatile experiment facility can conduct real-engineering system test which is designed to support the PCS development. In addition, this valuable experience in the design and manufacture of integrated experiment facility can provide important technical support and guidance for the China next generation advanced PWR as well as safety related system.
Determination of Gross Alpha and Gross Beta Activity Concentration in Bahe River Water Samples Within Xi'an Chanba Area
Authors: Qingxu Yao , Yonggang Huo , Peng Xu , Fengmei Yu , Ning Lv 
Published: October 2020

Abstract
As a screening procedure, gross alpha and gross beta activity have been developed to determine whether further analysis of water samples related to specific radionuclide is necessary. In China, the determination of gross alpha and gross beta in drinking water was generally based on the HJ standard method: HJ 898-2017, water quality — determination of gross alpha activity — thick source method, and HJ 899-2017, water quality — determination of gross beta activity — thick source method. In this study, 15 water samples from Bahe river in Chan Ba region of Xi’an in China, were pretreated and analyzed by BH1216-III low background alpha and beta scintillation counter.
The water samples were collected nearby residential area, wetland park, water conservation district and urban sewage treatment plant as well as other important locations which probably influence on the radioactivity level. The values of the activity concentrations of the gross alpha and beta measured in the water samples ranged from less than LD to 0.183Bq/L with a mean of 0.077Bq/L and 0.073–0.151Bq/L with a mean 0.102Bq/L respectively. All values of samples were lower than the limit level of 0.500Bq/L for gross alpha and 1.000Bq/L for gross beta, indicating that the radioactivity level in Bahe water between Chan Ba region of Xi’an is basically within the normal environmental background.
Technical Maturity Assessment of Risk-Informed Safety Analysis Tools
Authors: Choi Yong-Joon 
Published: October 2020

Abstract
Ensuring maximum safety while enhancing economic benefit is one of most important goal of In the of US Light Water Reactor Sustainability (LWRS) program. Optimization of the safety margins will provide best practice to achieve this goal which can also lead to cost reduction. Under the LWRS framework, the Risk-Informed Systems Analysis (RISA) Pathway has been focusing on the optimization of safety margin and minimization of uncertainties to ensure both safety and economics at the highest level. One of the important activities of the pathway is to deploy risk-informed analysis tools to related nuclear industry to support precise representation of safety margins and factors that contribute to cost and safety. The tools therefore need highest technical maturity so that industry can use immediately with strong credibility. The tools should have a capability to support risk-informed decision making for both probabilistic and deterministic elements of safety. The RISA Pathway, therefore, have been performing a comprehensive assessment of technical maturity and verification and validation (V&V) status of selected tools to improve adaptability to the industry. The technical maturity assessment includes three work scope: (a) define requirements based on risk-informed concept; (b) investigate and review development and V&V status for technical maturity assessment; and (c) identify technical gap and propose improvement to meet RISA toolkit requirements. The Requirement Traceability Matrix (RTM) concept was used to capture the requirements from user and developer of the project and/or software. The importance of each requirements was evaluated by Phenomena Identification and Ranking Technology (PIRT) which systematically gathers information and ranks the importance of the information. Finally, degree of the maturity was measured by Technology Readiness Level (TRL) for estimating the maturity of the technologies during the development and acquisition phase of certain technology. This paper summarizes development of assessment method and technical evaluation of multi-purpose probabilistic risk analysis tool RAVEN.
A Novel System for Automated Proper Use Identification of Personal Protective Equipment in Decommissioning Site of Fukushima Daiichi Nuclear Power Station
Authors: Chen Shi , Demachi Kazuyuki 
Published: October 2020

Abstract
Decommissioning of the Fukushima Daiichi nuclear power station is challenging due to industrial and chemical hazards as well as radiological ones. The decommissioning workers in these sites are indicated to wear Personal Protective Equipment (PPE) appropriately for radiation protection. In response to the difficulties of on-site PPE management in decommissioning site of Fukushima Daiichi NPS, this paper proposes the combination of deep learning-based object detection and individual detection using geometry relationships analysis to automatically facilitate the safety monitoring task of decommissioning workers to ensure hardhats, full-face masks and dust masks are correctly used for each zone. The experimental results demonstrate that the high precision, high recall and fast speed of the proposed approach can effectively detect decommissioning workers’ incorrect use of PPE.
Effects of Domain-Size on Large-Eddy Simulation for a Forced-Convection Flow in a Pipe With Circumferentially-Varying Heat Flux
Authors: Hattori Yasuo , Suto Hitoshi , Nakao Keisuke , Hasebe Yuma , Ishihara Shuji 
Published: October 2020

Abstract
We performed large-eddy simulations by using an open source CFD code, OpenFOAM to examine the performance of large-eddy simulations for examining turbulence heat-transfer processes of forced-convection in a pipe with non-homogeneous thermal boundary conditions; an accurate description of such processes is of practical interest of nuclear engineering. Special attention was paid to the domain size in the streamwise direction, which must be closely related to the turbulence processes with super structures. Three domain sizes were used: the size was varied from 5R to 100R, where R is the radius of the pipe. The turbulence intensities of temperature fluctuations near the heated surface strongly depended on the domain size. This was because that the turbulence intensities were closely related to large-scale fluid motions, the scale of which is much larger than 25R, and such large-scale fluid motions interfered in the dynamics of turbulence heat transfer processes near the heated surface. This indicates that the large-eddy simulations for estimating the turbulence heat-transfer rate in the pipe must require the large domain size in the streamwise direction with well resolved grids to capture turbulence eddies near the heated surface.
Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants
Authors: Hirano Sei , Hirasawa Daisuke , Kiyotoki Yoshihisa , Sakemura Keisuke , Sasaki Keiji , Fukuda Yuta 
Published: October 2020

Abstract
Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system.
Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements.
Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.
Modeling of Thermal Hydraulic Characteristics for a LBE-Cooled Fast Reactor Helical Coiled Type Steam Generator
Authors: Ding Xueyou , Wen Qinglong , Chen Zhiqiang , Ruan Shenhui , Cheng Cheng 
Published: October 2020

Abstract
Lead and Bismuth Eutectic (LBE) cooled fast reactor is attracting attention due to its advantages in safety. A thermal hydraulic analysis model is developed for an LBE cooled fast reactor helical coiled type steam generator, which could be used for predicting thermal hydraulic parameters distribution along the length of the tube. Based on two fluid model, the mathematical heat transfer model includes six regions: single water and vapor, subcooled boiling, saturated boiling, transition boiling and film boiling. In order to describe the heat and mass transfer between phases, the interphase model is presented. The steam generator is simplified with single tube concept and all the flow variables are evaluated at each point of the grid in which the domain is discretized. Full load condition and a postulated scenario, such a flow rate step changed in secondary side, is calculated in this study. The results showed that the changing process of the thermal hydraulic parameters of helically coiled steam generator conforms to the qualitative mechanism analysis results of thermal hydraulic analysis.
NICSim: Nuclear Instrumentation and Control Simulation for Evaluating Response to Cyber-Attacks
Authors: El Genk Mohamed S. , Schriener Timothy , Altamimi Ragai , Hahn Andrew , Lamb Christopher , Fasano Raymond 
Published: October 2020

Abstract
Digital Instrumentation and Control (I&C) systems in critical energy infrastructure, including nuclear power plants, raise cybersecurity concerns. Cyber-attack campaigns have targeted digital Programmable Logic Controllers (PLCs) used for monitoring and autonomous control. This paper describes the Nuclear Instrumentation and Control Simulation (NICSim) platform for emulating PLCs and investigating potential vulnerabilities of the I&C systems in nuclear power plants. It is being developed at the University of New Mexico’s Institute for Space and Nuclear Power Studies (UNM-ISNPS), in collaboration with Sandia National Laboratories (SNL), with high fidelity emulytics and modeling capabilities of a physics-based, dynamic model of a PWR nuclear power plant. The NICSim platform would be linked to the SCEPTRE framework at SNL to emulate the response of the plant digital I&C systems during nominal operation and while under cyber-attack.
Development of a Selection Tool for Choosing Decontamination Technology for Canadian Applications
Authors: Khurmi Raj , Carlisle Richard , Harvel Glenn 
Published: October 2020

Abstract
Decontamination is a major activity in decommissioning of Nuclear Power Plants. In efforts to reduce the overall volume of nuclear waste, retrieve reusable materials, and reduce the environmental impact, many different technologies have been developed/used in prior decommissioning projects and many more are being developed. However due to the amount of technologies available and the specific use cases for each, the ability to choose an appropriate and optimal technology is a challenge. An approach was adopted to develop a tool to assist in selection of decontamination technologies appropriate for Canadian Applications. The first step is the creation of a database to compile information of the different decontamination methods currently available in one location. The next step was the development of a software program to provide a search optimization for the database based on a set of initial user conditions. The program considers a radio-isotopic breakdown of a component as identified by the user and compares its concentration (Bq/g) to regulation limits set by the Canadian Nuclear Safety Commission (CSNC) for Unconditional Clearance Levels. Then, by using the CNSC guidelines, it will determine if the component is under Unconditional Levels or not. If the component is not, the code will calculate the minimum cumulative Decontamination Factor (DFR) required to make the component compliant with unconditional requirements. The software allows for users to plan their decontamination roadmap at a present state as well as a future state where natural decay opens up the ability for a wider range of decontamination technologies and for a combination of multiple components to use a given decontamination technique.
Structure Topology Optimization Design and Shock Resistance Study on Nuclear Power Safety DCS Cast Aluminum Cabinet
Authors: Dongwei Wang , Mingxing Liu , Xiao Wu , Hao Yan , Zhiqiang Wu 
Published: October 2020

Abstract
Offshore floating nuclear power plant (FNPP) is characterized by its small and mobility, which is not only able to provide safe and efficient electric energy to remote islands, but to the oil and gas platforms. The safety digital control system (DCS) cabinet, as a carrier for the electronic devices, plays a significant role in ensuring the normal operation of the nuclear power plant. To satisfy the requirements of cabinet used in the sea environment, such as well rigidity, shock load resistance, good seal and corrosion resistance, etc, more and more attention is focused on the cast aluminum cabinet. However, the cast aluminum structure may cause larger weight of cabinet, which inevitability affects the mobility of cabinet, and increases the carried load of ship as well. Therefore, seeking for an effective approach to design a light weight cast aluminum cabinet for the offshore FNPP is definitely necessary.
In this work, a frame of cast aluminum cabinet with lightweight is obtained successfully via structure topology optimization design, it is found that the weight of the frame can be reduced to 50% after optimization iterations. Subsequently, the natural frequency of the optimized cast aluminum cabinet is calculated by using ABAQUS, it is seen that the first mode frequency of the frame is beyond 30 Hz, which can meet the basic stiffness requirement. Accordingly, dynamic design analysis method (DDAM) is performed to verify the ability of the optimized cast aluminum cabinet in resisting sudden shock load, and the shock response characteristics of the cabinet are determined. Numerical results support that the optimized frame of cabinet possesses good resistance to high level shock. However, for the assembled cast aluminum cabinet, the vertical shock circumstance turns out to be the most critical condition, high stress and deformation regions occurs at the bracket and column. Reinforcements are proposed to make the bracket stiffer in this shock loading condition.