bluuebunny/crossref_metadata_2025_split huggingface dataset:
Application of High-Precision Assembly Technology for Large Structures by Laser Beam Welding
Authors: Nishiyama Tomoyuki , Kagawa Takashi , Tsubota Shuho , Kimura Masahiro 
Published: October 2021

Abstract
Toroidal Field coils (TF coils), which produce strong magnetic field in ITER, are used in order to confine the high temperature and high density plasmas stably for fusion reaction. The conductor made of Nb3Sn strands transform a superconducting conductor by heat treatment. The superconducting conductor is inserted into the D-shaped Radial Plate (RP), which have the structure with spiral grooves on both side for it, and then, insulating tapes are wound step by step around the superconducting conductor in RP as taking up the superconducting conductor. After insertion of the insulation-treated conductor into RP, Cover Plates (CPs) are put on RP spiral grooves and seal welded. After welding of CPs on RP of both sides, that the D-shaped Double Pancake (DP) is completed. The accuracy affects the TF coil performance and the deviation from ideal shapes and locations of the coils generates so called ‘error field’. In order to start up and confine the plasma stably, the allowable amount of the error field is about a few Gauss (Gauss : 1 × 10-4 T)[1]. Therefore, the dimensional accuracy of the 1/10000 order such as the profile within ±3mm for the circumference length and the flatness within ±3mm in the whole area of the D-shaped DP (14m × 9m) of approx. 37 m is required by ITER design. In addition, the TF coil is a huge component (with a weigh of about 300 ton) to withstand the high electromagnetic force of hundreds of MN.
It is an essential requirement for ITER design to achieve the accuracy of more than 1/10000 even though machining after welding is not allowed. Laser beam welding technology for large structure with thick plate of Mitsubishi Heavy Industries was applied the TF coils to minimize and avoid large welding deformation in the conventional welding process such as TIG welding to meet the ITER requirements.
A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations
Authors: Leppänen Jaakko , Valtavirta Ville , Tuominen Riku , Rintala Antti , Lauranto Unna 
Published: October 2021

Abstract
The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.
Application of Best-Estimate Plus Uncertainty Analysis Method in Nuclear Safety Evaluation
Authors: Tian Xinlu , Jing Jianping , Zhuang Shaoxin , Chen Haiying 
Published: October 2021

Abstract
Most of the computer codes for nuclear power plants safety analysis were developed based on conservative methods, which would lead to unnecessary excessive margins and more difficulty for nuclear power plants design and improvement. Using best-estimate plus uncertainty analysis method may reduce or eliminate these unnecessary limits. The NRC had used the code as a basis for safety reviews of the AP series nuclear power plants, driving the development of the best-estimate plus uncertainty analysis methods (BEPU) in the independent audit calculation. It had become an international trend to use BEPU in accident analysis and safety review of nuclear power plants. At present, the calculation method of the best-estimate in the world is used to better understand the accident sequence in the water-cooled reactor, and uncertainty assessment is the necessary supplement to the best-estimate method. The research and application of BEPU method are becoming more and more extensive. Best-estimate method was combined with best-estimation program and the initial and boundary conditions. It accurately described the current safety margin. It was required to consider each impact analysis of the available information and data, including the best-estimate model, the uncertainty calculated in the transient state, the input and the uncertainty of input parameters. Compared with the conservative evaluation method, the BEPU analysis method defined the gap between the calculated result and the real value through the uncertainty analysis, and made a more reasonable evaluation of the safety margin. In the process of nuclear power plant safety review, National Nuclear Safety Administration (NNSA) had adopted the best-estimate plus uncertainty analysis method to review the LOCA accident. In this paper, the four best-estimate plus uncertainty analysis methods are introduced, the source of uncertainty and uncertainty statistic methods are described. Then, the application of best-estimate plus uncertainty analysis method in nuclear safety evaluation are discussed.
Research on Remaining Useful Lifetime Prediction Methods of Main Transformer in Nuclear Power Station
Authors: Li Zikang , Peng Minjun 
Published: October 2021

Abstract
The main transformer of the NPP is the key equipment connects with public power grid and it is also the important electrical equipment to provide power for the power plant electrical equipment during the shutdown of the generator set. In consequence, it’s failure will cause huge consequences and economic losses. Therefore, the research of reaming life prediction technology for nuclear power plants(NPP) is of great significance to the economic operation and safety of NPP.
In this paper, according to DL/T984-2005 and other related literature, it is determined that the aging condition of solid insulating fiber material is the main factor to determine the life prediction of transformer. The main aging mechanism of that is summarized, and the effect of temperature is emphatic introduced. In general, the ambient temperature and load curve are the important factors affecting the aging of transformer insulation, furthermore affecting the life of transformer. Therefore, based on GB/T15164-94 recommended ambient temperature and transformer load, this paper calculates the transformer life loss model.
On this basis, this paper introduces the basic particle filter (PF) and Kalman filter (KF) algorithm flow. Based on the physical formula, a data model is developed to estimate the transformer life loss. Besides, the results are compared and analyzed. The study finally found that there were no significant differences in the accuracy of the predictions. However, considering the fact that the aging of transformer is a monotonic decreasing process, the particle filter based on degradation rate has a slight advantage in life prediction.
Three-Dimensional Numerical Simulation on Transient Natural Circulation Characteristics of DRACS in PLANDTL-DHX Experimental Device
Authors: Chen Zijia , Lu Daogang , Guo Jinsong , Zhang Yuhao 
Published: October 2021

Abstract
The pool-type sodium-cooled fast reactor adopts a new type of decay heat removal method to discharge the core heat to ensure the safety and cool of the nuclear power plant in a long time under the station blackout accident. The natural circulation and forced circulation will lead to different thermal and hydraulic characteristics. In addition, the natural circulation produces different hydrothermal phenomena on the secondary side by operating Direct Reactor Air Cooling System (DRACS). The Plant Dynamics Test Loop (PLANDTL-DHX) experimental device can simulate the flow and heat transfer phenomenon in the core. In the present work, a “modular modeling and integrated calculation method” was developed to conduct three-dimensional numerical modeling and calculation for the PLANDTL-DHX experimental device. Meanwhile, the steady-state verification of PLANDTL-DHX was carried out and compared with the experimental data. On this basis, the transient operating conditions of DRACS were calculated, and the temperature, flow distribution and transient variation characteristics in the key positions such as center subassembly, upper plenum and Intermediate Heat Exchanger (IHX) were preliminarily obtained. Based on the experimental results, the numerical calculation is verified.
ARKADIA – for the Innovation of Advanced Nuclear Reactor Design
Authors: Ohshima Hiroyuki , Asayama Tai , Furukawa Tomohiro , Tanaka Masaaki , Takata Takashi , Enuma Yasuhiro 
Published: October 2021

Abstract
This paper describes the outline and development plan for “Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA),” which the Japan Atomic Energy Agency has begun developing for the purpose of transforming advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize artificial intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for any challenge that could arise in the design and operation of a nuclear plant, including optimization of safety equipment as well as structures, systems, and components. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of a sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity that is technology inclusive and applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.
A High-Temperature Gas-Cooled Reactor Simulation System and Its Application Based on VPOWER Platform
Authors: Xie Biheng , Ji Chengzhi , Guo Xiaoyu , Han Wenbin , Hao Yisheng , Chen Junyi , Huang Shanfang , Wang Kan , Wei Hongbin , Liang Yanming 
Published: October 2021

Abstract
Tsinghua University and Neoswise Science & Technology Co. Ltd jointly developed the HTGR accident simulation system (EHTR) relying on the VPOWER platform. Based on Tsinghua University’s research into HTGR, the simulation system can provide users with a reliable HTGR control system that supports the functions of accident insertion, accident evolution, and accident analysis. At the same time, a friendly logical configuration interface is also provided. In nuclear engineering education, this system can help students deepen their understanding of the operation with advanced reactor equipment, which will help them enter the nuclear industry in the future. In this paper, EHTR is used to simulate the loss of feedwater accident. The changes in nuclear power, thermal power, temperature, and mass flow rate of the steam generator are analyzed. The simulation results are in good agreement with the design values, which shows the reliability of EHTR. Finally, the hybrid interactive teaching method combined with virtual simulation experiments and online teaching is introduced and applied in nuclear engineering courses. The teaching feedback shows that through the hybrid interactive teaching, the students have a better understanding of reactor operation knowledge, and stimulate their enthusiasm in the nuclear industry.
Analysis of IP200 Severe Accident Process Response to SBO and Emergency Power Failure
Authors: Zheng Zhenhang , Peng Minjun , Yu Hao , Yang Yang 
Published: October 2021

Abstract
Advanced SMRs such as the integrated pressurized water reactor IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In this thesis, the IP200 severe accident induced by the SBO and emergency power failure was modeled and analyzed using RELAP5 / SCDAP / MOD3.4 code. Based on the steady state calculation, which agrees well with designed values, the SBO accident for transient calculation is carried out. First, the case of the SBO accident without the passive core cooling system was calculated. The progression and scenario in the RPV was simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. Then, mitigation measures PRHRS and CMT were put in at four different time points when the core is began to uncovered, the core is completely uncovered, hydrogen is began to produced, and the molten pool is formed. The results show that putting in mitigation measures before the accident progresses to the point where the core starts to produce hydrogen can ensure that the core does not melt and avoid hydrogen risk.
Research and Application of Different Seismic Analysis Methods in Nuclear Power Equipment
Authors: Huang Xuan , Xiong Furui , Liu Shuai , Qi Huanhuan , Huang Qian , Zhang Ke 
Published: October 2021

Abstract
The integrity and reliability of nuclear power equipment are directly related to nuclear safety. According to ASME safety classification standard, the supports of pressure vessel, main pump and main cooler in reactor coolant system belong to nuclear safety class I and seismic class I. In order to verify the safety of the supports, according to the relevant regulations, it is necessary to carry out seismic analysis. The response spectrum method is based on the modal results, which can more accurately reflect the dynamic characteristics of the structure, so it is widely used in nuclear power engineering. Response spectrum method includes single point response spectrum method and multi-point response spectrum method. The single point response spectrum method uses the same seismic spectrum as the calculation input, which is suitable for the equipment with single support and small span; while the multipoint response spectrum method inputs different seismic spectra at different constraint points, which is suitable for equipment with multiple supports and large span. In this paper, the above two methods are used to carry out the seismic analysis of the main equipment support of the reactor coolant system, and the application of different seismic analysis methods in nuclear power equipment is studied.
A Numerical Study of Supersonic Film Cooling With Discrete Holes
Authors: Ni Hang , Peng Wei , Wang Jie , Zhu Yinhai , Jiang Peixue 
Published: October 2021

Abstract
Nuclear thermal propulsion is an effective propulsion technology that can meet the needs of deep space exploration. The nozzle of the nuclear thermal propulsion system endures high-temperature and high heat flux environment, while supersonic film cooling is an important cooling method to provide thermal protection. The present study numerically investigated the effect of the blowing ratio and the shock wave on supersonic film cooling with discrete holes. The results show that when the blowing ratio is large, a strong kidney vortex is formed and the high-temperature mainstream enters the cooling layer. Therefore, the film cooling effectiveness at the upstream region near the hole decreases rapidly at a large blowing ratio. For the middle and downstream regions, the increase of blowing ratio gives a larger coolant mass flow rate and thereby improves film cooling effectiveness. In general, film cooling effectiveness decreases with shock wave impingement. For a strong shock wave, the cooling stream shows strong lateral diffusion after the shock wave impinging region. As a result, the lateral average film cooling effectiveness increases first and then decreases. This phenomenon is more obvious at a larger blowing ratio.
Severe Accident Analysis of a Floating Nuclear Power Plant After Station Black Out Accident
Authors: Li Longze , Chao Fei , He Chuan , Tai Yun , Qiu Jinrong , Yao Shiwei , Yang Wen , Wang Jue , Hou Xiaofan 
Published: October 2021

Abstract
The Floating nuclear power plant (FNPP) is a kind of marine nuclear power plant which takes barge as carrier and is fixed in specific sea area by mooring device. A 100 MWt PWR type small modular reactor (SMR) with 2 loops is used in the FNPP. It is designed based on the standard and experience of the traditional onshore nuclear power plant. The engineering safety features in the plant are somewhat different from those in the traditional plants. The special residual heat removal system, the passive residual heat removal, the square steel containment and containment suppression system are designed in the plant. Three severe accident mitigation measures, i.e., the pressurizer (PZR) relief extension, the external vessel reactor cooling (EVRC) and the Passive autocatalystic recombiners (PAR) are set up to actually eliminate the large release of radioactive products. The severe accident induced by station black out (SBO) for the SMR in FNPP is analyzed with MELCOR code in the present work. The MELCOR model were developed for the main systems in the SMR, including the primary and secondary system, the engineering safety features, the containment system, and the severe accident mitigation measures. The severe accident sequence as well as the key phenomena during the accident are analyzed. The simulation results indicated that the core exposed and heat up for the coolant leakage via the pressurizer safety valve, and finally melt with the high pressure and lower pressure injection failure. However, the reactor pressure vessel (RPV) maintained integrity with the mitigation measure, i.e., the external vessel core cooling. The containment also maintained integrity, which prevented the large release of radioactive products to the other cabins and the environment. The work is useful in gaining an insight into the detailed severe accident process involved, and could be used to identify appropriate severe accident management strategies and countermeasures for the FNPP.
100-Gigawatt-Hour Crushed-Rock Heat Storage for Variable Electricity and Heat With Base-Load Reactor Operations
Authors: Forsberg Charles 
Published: October 2021

Abstract
A 100-gigawatt hour (GWh) crushed-rock heat storage system using oil or nitrate salts is proposed to enable base-load 1000-MWe nuclear plants to provide variable electricity and heat to maximize plant revenue. Such a heat storage system has the same capabilities as a large pumped-hydro facility and can provide hourly-to-weekly heat storage. The capital cost goal is $2–4/kWh of heat — more than an order of magnitude below electricity storage technologies (batteries, pumped storage, etc.). Oil (LWR) or nitrate salts (higher-temperature reactors) transfer heat from the reactor to storage and from storage to the non-nuclear power block. The non-nuclear power block output is several times base-load nuclear capacity.
The insulated heat storage structure would have a crushed rock bed 20 m by 250 m by 250 m. The large dimensions minimize the surface area (steel and insulation) and cost per cubic meter of crushed-rock storage. Crushed rock is the lowest-cost heat storage material. Heat is transferred from the reactor to crushed rock by spraying on sequential sections of crushed rock (1) hot oil or (2) molten nitrate salt. The liquid flows by gravity to the drain pan at the bottom of the crushed rock and is returned to the reactor to be reheated. Heat is recovered by spraying cold fluid on the top of the hot crushed rock with hot liquid recovered from the drain pan at the bottom going to the power cycle. The oil or nitrate salt is used for heat transfer — not for heat storage, thus, minimize the quantity and cost of the heat transfer fluid.
Nuclear Air-Brayton Combined Cycles Using Electrically-Heated Conductive Firebrick Heat Storage and Hydrogen for Peak Power
Authors: Forsberg Charles , Stack Daniel , McDaniel Patrick 
Published: October 2021

Abstract
A low-carbon world needs a replacement for the combined-cycle gas turbine to provide variable heat and electricity. We describe a nuclear combined cycle gas turbine (NACC) with a thermodynamic topping cycle where the high-temperature heat is provided by an electrically-heated thermal energy storage system or hydrogen. During base-load operations the NACC compressed air is heated up to 700°C with heat from a high-temperature nuclear reactor to produce electricity. At times of low electricity prices, electricity from the NACC and the grid is converted into high-temperature stored heat in the form of hot firebrick. At times of high electricity prices compressed air is heated up to 700°C by the reactor and then further heated by sending the compressed air through hot firebrick to heat the air up to 1427°C before going through the turbine and heat recovery steam generator to produce added peak electricity. The round-trip electricity-to-heat-storage-to-electricity efficiency is above 70%; that is, if one unit of electricity is sent to heat storage, more than 0.7 units of peak electricity are returned. The enabling technology is a new doped electrically-conductive firebrick that enables conversion of electricity to very-high-temperature stored heat. Stored heat addresses daily variations in demand. Hydrogen can be used for long-term seasonal energy storage but at lower peak electricity efficiency because the efficiency of the electricity-to-hydrogen conversion step is near 75% versus near 100% for electricity-to-heat.
Research on Radioactive Consequence Limits for SSG-30
Authors: Zhao Danni , Pang Zongzhu , Li Ming , Liu Yu 
Published: October 2021

Abstract
After the release of new nuclear safety requirements SSR-2 / 1 (the latest version is 2016 Edition) in 2012, IAEA issued a new guideline SSG-30 “safety classification of structures, systems and components of nuclear power plants” in 2014 proposing a new safety classification method. Subsequently, IAEA published TECDOC-1787 “Application of Safety Classification of Structures, Systems and Components of Nuclear Power Plants” in 2016, which further explained the application of SSG-30. The safety classification method of SSG-30 based on the safety functions and design provisions. An SSC implemented as a design provision should, however, be classified directly because the significance of its postulated failure fully defines its safety class without any need for detailed analysis of the category of the associated safety function. The severity of consequences is divided into three levels (high, medium and low). The high, medium and low consequences are defined by the radiation dose criteria for the public and staff in different plant conditions. In order to classify the structures, systems and components of nuclear power plant, the specific meaning of the definition of high, medium and low “consequences” should be clarified firstly. In this paper, the radioactive dose criteria for the public and staff under different working conditions of nuclear power plant are studied and suggestions are put forward to make SSG-30 more operable when used in nuclear power plant SSCs classification.