bluuebunny/crossref_metadata_2025 huggingface dataset:
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

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

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

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

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

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

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

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

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

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

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

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

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.
Research on Nuclear Safety Video Display Unit Technology Based on Digital Twin
Authors: Wu Yanqun , Zhao Yang , Liu Mingxing , Jiang Wei , Chen Qi , Wang Shun , Yan Hao
Published: October 2020

The nuclear safety video display unit (SVDU) in the nuclear safety level DCS implements functions such as displaying and warning of reactor safety related parameters and sending safety control instructions, which belongs to safety level 1 equipment. As a high-cost complex safety level display device, due to its high functional complexity, high security and reliability requirements, and special usage scenarios, its design, research, and function verification have always encountered problems such as low intelligence and low digitization, resulting in slow development process, complicated product function verification, and inconvenient use and training, etc. After the SVDU is put into practical use, continuous analysis of its stability, reliability, and safety, and its health status monitoring and maintenance are also difficult technical problems.

Based on the five-dimensional digital twin model as a design benchmark, a digital twin-based design method for SVDU is proposed. Taking the SVDU in a nuclear safety level DCS (NASPIC) as the object, the digital twin technology is adopted to model the physical objects such as display unit, human-machine interface unit, storage unit and communication network unit, and the high-speed industrial Ethernet network is used to map and interconnect the data between the components, so as to realize the physical fusion, model fusion and data fusion of the real SVDU and the virtual SVDU. With the help of data feedback from safety level DCS, the data, symbol display and control process of SVDU can be visualized and analyzed in virtual environment, and the real-time control function verification, fault early warning and auxiliary decision-making can be carried out, which improves the visibility, accessibility, operability and predictability of real SVDU display and control process. The real-time data, historical data, fault self-diagnosis data, and expert experience of real SVDU and virtual SVDU are incorporated into the twin data pool to reproduce the variable-speed replay of the historical operation process of SVDU and realize the post-accident condition analysis with multiple spatiotemporal dimensions; the record and analysis of self-fault diagnosis data provides the possibility of stability analysis services such as clock, power supply, human-machine input, and storage equipment, etc.

Digital twin-based SVDU technology can ensure rapid development and iteration of products under the requirements of complex functions and high safety and reliability, and truly reproduce the display and control effects of SVDU, while meeting users’ multiple application scenarios and data service requirements.
Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions
Authors: Yin Chunyu
Published: October 2020

SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.
Use of Micro-Gravity Sensors for External Fluid Level Monitoring in Waste and Nuclear Related Applications
Authors: Ridgway Jeff , Slater Bryant
Published: October 2020

There are a number of applications in nuclear energy and hazardous waste disposal that require monitoring of fluids under extreme environments, including high levels of temperature, pressure, toxicity and radioactivity. Many of these applications will benefit from a monitoring technique that is external and non-invasive. Currently the sensors used are invasive, must reside inside the pressurized vessels and must penetrate the vessel walls, which can create a weakness in the vessel. Additionally, instruments that are used inside such containers must be exceptionally hardened to the environment. Information Systems Laboratories (ISL) has developed an external mass (gravimetric) measuring technique for monitoring nuclear coolant in Small Modular Reactors (SMRs), which will also work for measuring fluid levels in waste tanks, that avoids the problems inherent in invasive sensors. It utilizes a COTS gravitational sensor of unprecedented accuracy, leveraged via proper sensor placement geometry, to detect fluid changes of small amplitude from an outside position, obviating the need to penetrate the vessel. The technique is called Gravisense™.

ISL has proven via simulation and experiment that this concept can be usefully applied to monitoring fluid levels in both nuclear reactors and large waste tanks. Numerical simulation algorithms were developed to calculate the gravity effect of small changes in water level, which were verified by experiments at the NIST Physical Simulator facility at the Oregon State University. The measured ultra-low noise levels of the superconducting gravimeter type which utilizes a Niobium sphere suspended in a magnetic field to attain its phenomenal accuracy, demonstrated that fluid levels in SMRs can be measured at least to within 3 cm. Furthermore, the method can distinguish between a contained leak (from reactor to containment vessel) from an external leak (from reactor to outside of containment).

Additionally, simulations of waste canisters that hold spent fuel rods show that the fluid level measuring accuracy can potentially do better than 1 cm accuracy by measuring from below the vessel, and judicious placement of sensors on top of large waste tanks can potentially achieve a very impressive 2 mm measurement accuracy.

These encouraging results prove that the Gravisense™ technique for fluid determination can be very useful in nuclear energy generation, testing, and research, as well as in waste monitoring situations that are difficult to monitor via traditional sensing technology. We believe that the next step should be to test the technique on canisters of the type that are currently storing waste in various DOE locations.
A Numerical Simulation Method for Core Internals Behavior in Severe Accident Conditions: Chemical Reaction Analyses in Core Structures by JUPITER
Authors: Yamashita Susumu , Kino Chiaki , Yoshida Hiroyuki
Published: October 2020

In order to contribute the improvement of estimation accuracy for severe accident code such as SAMPSON, we have developed the chemical reaction model such as eutectic reaction and oxidation in micro scale, e.g., B4C-SUS in the control rod blade and UO2-Zry in fuel rods, and implemented them to the computational fluid dynamics code named JUPITER. And we try to develop the coupled analysis frame work using SAMPSON and JUPITER to decrease uncertainty due to micro scale phenomena which cannot be calculate by severe accident analysis codes.

From the preliminary analysis in fuel rod heating analysis by JUPITER using SAMPSON output data, it was revealed that the implemented chemical reaction models work stably and obtain reasonable results.
Sensitivity Analysis for Release and Transport Behavior of Radionuclide of the 1F Unit-1 Accident Using SAMPSON
Authors: Kino Chiaki , Karasawa Hidetoshi , Uchida Shunsuke
Published: October 2020

To support the 1F decommissioning project, the Institute of Applied Energy has been analyzing the course of the accident using the SAMPSON code with an aim to grasp radionuclide distribution in 1F NPP. The present study has mainly focused on molybdenum release from a fuel pellet and chemical forms of iodine during the early phase of the accident because a release rate of semi-volatile FPs has various uncertainties. The present simulation set the detrimental factor of molybdenum used in FP release model as a sensitive parameter. The simulation results indicate that more molybdenum release causes more generation of hydrogen iodide and more deposition of iodine inside Reactor Pressure Vessel (RPV) because the adsorption velocity of hydrogen iodide is comparatively higher than deposition velocities of another FPs.
Development of Effective Momentum Model for Steam Injection Through Multi-Hole Spargers Using a Condensation Region Approach
Authors: Wang Xicheng , Grishchenko Dmitry , Kudinov Pavel
Published: October 2020

The Pressure Suppression Pool (PSP) in a Boiling Water Reactor (BWR) is served as a heat sink to prevent containment over-pressure. The steam can be injected through the multi-hole spargers. The development of thermal stratification where a thermocline with a large temperature gradient appears in the pool can lead to the higher pressure in the dry well compared with completely mixing pool conditions. Prediction of the thermal phenomenon in the pool is necessary for the support of system design and operation. Thus, the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed. The models can be applied to CFD code by using (i) source terms in the transport equations or (ii) using respective boundary conditions at the Steam Condensation Region (SCR). Previous validation against PPOOLEX and PANDA tests using the source terms approach faced challenges in momentum distribution. Therefore, a preliminary investigation of using the second method was performed. The encouraging results implied that it is possible to further develop this approach. The goal of this work is to further develop the EHS/EMS models for the steam injection through a multi-hole sparger through the SCR model (i.e approach (ii)) and to validate it against the experimental data obtained from PANDA HP5 tests. Modeling guidelines are proposed. The temperature evolutions and vertical velocity profiles of these tests are compared to the simulation results. The agreement suggests that this model can provide an adequate estimation of the pool behavior.
Analytical Study of Perforation Damage to Reinforced Concrete Slabs Subjected to Oblique Impact by Projectiles With Different Nose Shapes
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Li Yinsheng
Published: October 2020

Considerable research has been carried out to establish a rational assessment method for nuclear power plants against local damage caused by an accidental projectile impact. Most of the empirical formulas that have been proposed seek to quantitatively investigate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived based on impact tests performed perpendicular to the target structure, while only a few impact tests oblique to the target structure have been studied. This study aims to propose a new formula for evaluating local damage to RC structures caused by oblique impact based on experimental and simulation results. At present, we have validated an analytical method via comparison with experimental results and have conducted simulation analyses of oblique impact assessments on RC slabs using various projectiles with flat nose shapes. In this study, the same analytical method will be used to investigate perforation damage to RC slabs subjected to oblique impact by projectiles with hemispherical nose shapes. In this paper, the effects of projectiles’ nose shapes on perforation damage to RC slabs, the residual velocity of projectiles and the time history of energy transmission will be discussed.
Applicability of FLASH-CAT Model to Cable Tray Fire Modeling in Zone Model BRI2002
Authors: Shirai Koji , Tasaka Koji , Udagawa Toshiko
Published: October 2020

To clarify the heat and smoke propagation in multi-compartments under the spread of cable fire, a large-scale multi-compartment fire test (hereinafter the CFS-2 test) was performed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in France within the framework promoted by the Nuclear Energy Agency (NEA) in Organization for Economic Co-operation and Development (OECD) program PRISME2 (OECD/NEA, 2017). In the CFS-2 test, two rooms of a large-scale facility were adopted and these rooms have an identical volume (120 m3) enclosed with fire walls and were connected by a doorway (0.8 m in width and 2.17 m in height). As a fire source, five-layer cable trays (tray length of 2.4m, tray width of 0.45m and separation distance between trays of 0.3 m) with a fire-retardant PVC cable (77 kg) were used and ignited by a propane gas burner. The power level of the propane gas burner was set to around 80 kW. Moreover, all rooms were mechanically ventilated, and the renewal rate was 15 times per hour (3600 m3/h). During the fire test, the mass loss rate of fuel, gas and soot mass concentration, gas temperature, and etc. were measured. The measured peak values of the HRR, the mass loss rate and gas temperature were about 800 kW, 58 g/s and greater than 600 °C, respectively (Zavaleta, 2017).

As a fire model predicting fire characteristics in a compartment, a two-zone model, which divides the fire room into the hot smoke upper layer and lower layer consisting of cool fresh air, is widely used due to the advantages of the brevity of the calculation routine and the reliability of the calculation results. Among them, the BRI2 series, developed in Japan, is now reaching the current BRI2002 software (Wakamatsu, 2004) after several upgrades to improve the calculation precision. The Central Research Institute of Electric Power Industry (CRIEPI) introduced the cable tray fire source model based on the FLASH-CAT (Flame Spread over Horizontal Cable Trays) developed by National Institute of Standards and Technology (NIST) (McGrattan, 2012) into the zone code BRI2002. By comparing the numerical results with the experimental values measured during the CFS-2 test, the methodology for ignition time delay of each tray and horizontal flame propagation speed for each tray were discussed.
Optimization Study for Uncertainty Reduction of In-Pile CHF Experiments in TREAT
Authors: Seo Seokbin , Brown Nicholas R. , Armstrong Robert J. , Folsom Charles P. , Jensen Colby B.
Published: October 2020

Reactivity-initiated accidents (RIAs) are one of the postulated incidents that can threaten the operational safety of a nuclear reactor. During a RIA, a rapid increase of energy deposition in the fuel can lead to a departure from nucleate boiling (DNB) occurrence which refers to the point where a drastic decrease in heat transfer capabilities occurs and the surface heat flux exceeds the critical heat flux (CHF). Aiming to understand the fundamentals beneath CHF and to predict it, the Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL) is a unique facility that will be used to experimentally investigate the transient CHF under in-pile pool boiling condition. As part of a comprehensive effort to utilize TREAT for this project, this study analyzed the expected uncertainties in the experimental data by identifying the key inputs for the uncertainty in the temperature measurements and quantifying their priorities. The sensitivities of key inputs from neutronics modeling, the clad-to-coolant heat transfer, thermophysical properties of the tube, and coolant conditions were quantified using Sobol sensitivity analysis methods, and the significant effect of the occurrence of the CHF on the sensitivity of input was found.
Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident
Authors: Lin Linyu , Athe Paridhi , Rouxelin Pascal , Dinh Nam , Lane Jeffrey
Published: October 2020

In this work, a Nearly Autonomous Management and Control (NAMAC) system is designed to diagnose the reactor state and provide recommendations to the operator for maintaining the safety and performance of the reactor. A three layer-hierarchical workflow is suggested to guide the design and development of the NAMAC system. The three layers in this workflow corresponds to knowledge base, digital twin developmental layer (for different NAMAC functions), and NAMAC operational layer. Digital twin in NAMAC is described as knowledge acquisition system to support different autonomous control functions. Therefore, based on the knowledge base, a set of digital twin models is trained to determine the plant state, predict behavior of physical components or systems, and rank available control options. The trained digital twin models are assembled according to NAMAC operational workflow to support decision-making process in selecting the optimal control actions during an accident scenario.

To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II (EBR-II) during a single loss of flow accident. Training database for development of digital twin models is obtained by sampling the control parameters in the GOTHIC data generation engine. After the training and testing, the digital twins are assembled into a NAMAC system according to the operational workflow. This NAMAC system is coupled with the GOTHIC plant simulator, and a confusion matrix is generated to illustrate the accuracy and robustness of implemented NAMAC system. It is found that within the training databases, NAMAC can make reasonable recommendations with zero confusion rate. However, when the scenario is beyond the training cases, the confusion rate increases, especially when the scenarios are more severe. Therefore, a discrepancy checker is added to detect unexpected reactor states and alert operators for safety-minded actions.
Study on Seismic Isolation and Hi-Frequency Vibration Isolation Technology for Equipment in Nuclear Power Plant Using Aero Floating Technique
Authors: Takito Kiyotaka , Furuya Osamu , Kurabayashi Hiroshi , Sanpei Kunio
Published: October 2020

In Japan, most structures on the ground surface need seismic countermeasures because of frequently earthquakes. On the other hand, vibration isolation devices are applied to precision or important equipment in several facilities that dislikes vibration in order to reduce daily vibration. In general, vibration isolation devices are intended for high frequency and small amplitude range. However, it is difficult to cut off both vibration region caused by flying object collision and seismic motion with existing technologies. The authors propose insulation of equipment and vibration transmitted through the floor by floating equipment, and have. We have devised and built an air floating device that operates when a trigger input is applied to save the energy of this dynamically acting device. It was estimated by numerical calculation that the aero floating device keeps lifting stably in the condition with the air pressure in the auxiliary air chamber about 75 to 80 kPa. The performance specifications of the proposed device were verified from shaking table test.

As a result, the effect of reducing the maximum acceleration by about 1/5 against the seismic motion of El Centro NS, Taft NS, Tohoku NS, and Hachinohe EW was confirmed by floating the mass on the frame assuming the equipment. From the obtained power spectrum diagram (PSD) of the response acceleration, it was confirmed that all frequency components up to 25 Hz is reduced by using proposed aero floating base isolation device.
Experimental Study for Evaluation of Spreading Behavior of Free-Falling Molten Core With Floor Impingement
Authors: Ito Tomomasa , Watanabe Yuta , Yamamoto Yasunori , Sahboun Nassim , Miwa Shuichiro
Published: October 2020

Understanding the spreading behavior of the molten core is important for predicting the progress of severe accidents and for smooth decommissioning work. In this study, we performed molten metal drop experiments using Sn and Cu to obtain fundamental data and its uncertainties that can be used to verify the analysis model and to evaluate the effect of scaling parameters on the spreading and deposition behaviors. We obtained spreading area and deposition thickness of the metals after floor impingement. The standard deviation on average were 6.0% and 6.7% for the thickness and the area, respectively. In addition, the correlation with each dimensionless number was confirmed, and strong correlations were obtained between the deposition behavior and the Re number, and between the spreading behavior and the Pe number. We constructed a correlation equation using our experimental data as a function of Re number, We number, Fr number, Pe number, and dimensionless length (L/d). The errors for our experimental data were within 30%.
Fundamental Study on Decontamination Method Using Shot Blasting and Barrel Polishing for Radioactive Metal Waste
Authors: Yamamoto Sota , Takahashi Hideharu , Taniguchi Hayato , Kawashima Akihiko , Takahashi Hiroshi , Kikura Hiroshige
Published: October 2020

In the decommissioning of nuclear power plants (NPP), we can greatly reduce the amount of low-level waste (LLW) by decontamination process. Generally, chemical decontamination method that use chemical solutions to dissolve the deposited waste is major way of decontamination. But this method takes a lot of time and generates secondary waste. On the other hand, the mechanical decontamination method that projected particles such as metal to decontamination target and remove the deposited waste takes shorter time than chemical decontamination method usually. However, in the case of mechanical decontamination of the small diameter pipe to be used in the primary system of the nuclear reactor is difficult because we must divide it vertically into half, and these small diameter pipes are disposed as LLW. At the decommissioning of advanced thermal reactor Fugen (operation finished in 2003) in Japan, the new decontamination machine which compounded with the blasting machine which projected particles such as metal and the barrel polishing machine which is a machine that polishes like a washing machine using particles such as metal was produced experimentally for the pipes of such a small diameter. By using this machine, the radioactivity of contaminated undivided pipe is decreased, and we found that this machine can use as a decontamination machine. In this study, we operated the blasting and barrel grinder independently in order to elucidate the mechanism of the decontamination system. We placed flat metal simulators in each device and measured mass changes in each to compare which decontamination effect was the primary one. The results suggest that the decontamination effect of the blasting machine is higher than that of the barrel polishing machine. We also used a high-speed camera to observe the decontamination of the blasting system and to investigate the interaction between the pipe sample and the particles. It was suggested that the number of particles entering the inside of the pipe increases as the pipe faces the projection direction.
Calculation and Analysis of Steam Hammer in Main Steam Pipe in HPR1000
Authors: Yu Pei , Wang Jiaming , Ma Huiyun , Gu Haifeng , Yan Changqi
Published: October 2020

The steam hammer pressure is solved though the simplified calculation. PIPENET software is applied to model the nuclear island main steam system between the steam generator and the main steam header in HPR 1000. The transient module is used to simulate the occurrence and attenuation process of steam hammer. The maximum steam hammer pressure, the maximum steam hammer stress in the pipe system, when and where the load occurs are given. The influence of the straight pipe section length and valve closing time on the steam hammer effect is analyzed. With the other conditions unchanged, the steam hammer energy decreases as the straight pipe section shortens, or the valve closing time extends.
Study of spatiotemporal liquid dynamics in a vibrating vocal fold by using a self-oscillating poroelastic model
Authors: Scholp Austin , Jeddeloh Caroline , Tao Chao , Liu Xiaojun , Dailey Seth H. , Jiang Jack J.
Published: October 2020

The main purpose of this study is to investigate the spatiotemporal interstitial fluid dynamics in a vibrating vocal fold. A self-oscillating poroelastic model is proposed to study the liquid dynamics in the vibrating vocal folds by treating the vocal fold tissue as a transversally isotropic, fluid-saturated, porous material. Rich spatiotemporal liquid dynamics have been found. Specifically, in the vertical direction, the liquid is transported from the inferior side to the superior side due to the propagation of the mucosal wave. In the longitudinal direction, the liquid accumulates at the anterior-posterior midpoint. However, the contact between the two vocal folds forces the accumulated liquid out laterally in a very short time span. These findings could be helpful for exploring etiology of some laryngeal pathologies, optimizing laryngeal disease treatment, and understanding hemodynamics in the vocal folds.
Greater Semantic Memory Activation After Exercise Training Cessation in Older Endurance-Trained Athletes
Authors: Won Junyeon , Alfini Alfonso J. , Weiss Lauren R. , Hagberg James M. , Carson Smith J.
Published: October 2020

Purpose: To examine the effects of a 10-day exercise-training cessation on semantic memory functional activation in older distance runners. Methods: Ten master runners (62.6 ± 7.0 years) with a long-term endurance-training history (29.0 ± 6.0 years) underwent a 10-day training cessation. Before and immediately after the training cessation, semantic memory activation was measured during the famous name recognition task, using functional magnetic resonance imaging. Results: The 10-day training cessation resulted in greater semantic memory activation in three brain regions, including the left inferior frontal gyrus, parahippocampal gyrus, and inferior semilunar lobule. The 10-day training cessation did not significantly alter famous name recognition task performance. Conclusions: The findings demonstrate that even a relatively short period without exercise training alters the functional activation patterns of semantic memory–related neural networks. Increased semantic memory activation after training cessation may indicate reduced neural efficiency during successful memory retrieval.