bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on the Water Intake of Heat Trap Blocked by Marine Life in Nuclear Power Plant
Authors: Zhang Zhenhua , Zhang Bo , Zheng Chaoying 
Published: November 2022

Abstract
In recent years, the operational events that the water intake of heat trap was blocked by marine life in nuclear power plants have occurred repeatedly. Since 2000, there have been more than 200 shutdown events of biological (foreign body) blockage resulting in failure of cooling water system in the United States, the United Kingdom, Sweden, Arabia, India, Israel, Japan, South Korea and China. It has caused major safety risks and brought huge economic losses and has attracted wide attention from various governments, related industries and research institutions. It can be seen that water intake blockage event is a safety problem for all kinds of power plants to face. This paper introduces the typical design of water intake and analyses the potential effects of a large number of marine life into the heat trap water intake from the perspective of circulating water system (CRF) jump pump and essential service water system (SEC) loss of function. Based on the characteristics of the high frequency of such operation events and they may cause major safety risks, the reactor shutdown response time in different conditions is simulated. When a large number of marine organisms flood into the water intake of heat trap, the response of the operator on the potential impact of the unit condition and the operational response plan are analyzed. The impact of the unplanned changes in operating status caused by the blockage of the water intake is explained. Ultimately, based on the analysis as described above, suggestions on the design improvement of the heat trap water intake and seawater filtration system in the nuclear power plant are presented. At the same time, suggestions are put forward on the supervision analysis, dynamic monitoring of the surrounding environment of the water intake, establishing a rapid decision-making mechanism, improving the cold source emergency response plan and increasing the operation control strategy.
Research on Optimization of Reload Startup Physics Tests in Sanmen Nuclear Power Plant
Authors: Shi Jianfeng , Yang Qingxiang , Wang Lihua , Chen Lijiang , Li Ang 
Published: November 2022

Abstract
During reload startup of a nuclear power plant (NPP), physics tests are performed to verify the consistency of core neutronic characteristics with design expectations. Physics tests time during power ascension directly affects the power rise rate and economy of units operation. The paper studied the physics tests process during power ascension of reload startup in Sanmen NPP, and proposed an optimization scheme for Sanmen reload startup physics tests based on test data analysis and simulation. On the premise of meeting the requirements of current guidelines and standards, the scheme optimizes the current test method from the aspects of test contents andtest initial conditions, which can effectively reduce risk to core operation, shorten test time and improve economy of units operation. In terms of test content, it was demonstrated that the influence of detector leakage current on the power distribution measurement can be ignored, which can reduce the times of leakage current measurement in the process of power rise. In terms of the test initial conditions, the influence of xenon on the test process and results was analyzed, so as to relax the initial conditions of xenon equilibrium and shorten the total power rise time from about 4 days to about 2 days. Standard compliance was also evaluated, and the optimizations are in accordance with the requirements in current standards.
Numerical Investigation on Transport Characteristics of Hydrogen-Air-Steam Mixture Induced by Condensation in the Single Compartment
Authors: Liu Feng , Sun Zhongning , Bian Haozhi , Ding Ming 
Published: November 2022

Abstract
Under the severe accident of the nuclear power plant, hydrogen-steam mixture gas will be released into the containment compartments. In this process, there is a strong coupling relationship between mixture gas diffusion and steam condensation on the compartment wall surface. Especially for local compartments with poor connectivity, steam condensation is likely to aggravate the risk of hydrogen accumulation and combustion. This coupling phenomenon, which has not been systematically studied in previous numerical simulations or experiments, is especially complex and needs to be further studied. Moreover, the applicability of mechanistic numerical models for predicting condensation-induced transient gas transport processes also should be further confirmed. In this study, the mechanistic numerical model was evaluated through the transient hydrogen mixing experiment of the THAI facility. Based on the validated model, the transient mass transfer process of the mixture gas induced by condensation in a typical single compartment during the steam-hydrogen release process is studied. Furthermore, the influence of release condition of gas source term, wall sub-cooling on the hydrogen accumulation phenomenon in the compartment was discussed. The conclusion is helpful to further understand the transport behavior of hydrogen-air-steam mixture gas in the local single compartment under the severe accident.
Simulation of MSLB+SBO Nuclear Accident Drill Sequence of 3rd-Generation Passive Nuclear Power Plant Based on Severe Accident Simulator
Authors: Yang Yongyong , Huang Xiong , Wei Wei , Tao Yinyong , Xie Zhengquan , Ma Guoyang 
Published: November 2022

Abstract
The full scope simulator (FSS) is very important for operator training, but the related SA training can’t be conducted for lack of the severe accident models, so it is urgent to expand the severe accident model on the FSS in nuclear power plant (NPP). The full scope simulator of 3rd-generation passive NPP can perform more abundant and complete accident scenes for operator training and accident drills through coupling MAAP5 models, namely severe accident simulator (SA simulator). In this paper, the severe models (such as core, primary system, containment), the key technologies of coupling MAAP5 software with FSS are described. Then taking the MSLB+SBO as the initial hypothetical condition to simulate the whole process of the severe accident emergency drill scenario in 3rd-generation NPP and make a preliminary sensitivity analysis on the mitigation effect of the mitigation measure for RCS injection flow in the accident drill scenario. The results show that the whole severe accident drill process can be very simulated, the SA simulator can not only provide an effective mean for the simulation of NPP severe accident drill scenarios, but also can be used as an auxiliary tool for the verification of severe accident management guidelines (SAMG) and the optimization of drill scenarios in NPP, which provide some technical support for the safe and stable operation of nuclear power plant.
Research on Quantitative Assessment Method of Security Guard Capacity in Nuclear Power Plants
Authors: Liu Xiaolin , Meng Wei , Li Jianwen , Zhang Jinfei 
Published: November 2022

Abstract
Under current circumstances, the daily training of the security team in nuclear power plant is carried out by the security company providing security services according to the requirements of the power plant security department, followed by manual assessment and authorization among the guards. There are several problems in this process that can be improved: evaluation is highly subjective, difficult to quantify, and difficult to standardize. Based on the principles of objectivity, unity, comprehensiveness and extendibility, this paper proposes an intelligent assessment method for nuclear power plant security guard capability. Firstly, it decomposes single task into core abilities which are aggregated to an ability library based on their corresponding tasks. Secondly, all task scenarios are made into VR test scenarios with core abilities as the test points, and aggregated into test banks. Then use the VR assessment system for automatically issuing, collecting and scoring test papers. Finally, a comprehensive assessment of the tested guard will be released with the combination of the fuzzy comprehensive assessment method. This method not only can increase objectivity and scientificity of assessment but also has extendibility to new job requirements to meet dynamic changes.
Optimization for Rapid Dose Calculation of the RIMPUFF Model and Its Evaluation Against Belgian Field Experiment
Authors: Yang Li , Li Xinpeng , Chen Yixue , Wang Boxin , Fang Sheng 
Published: November 2022

Abstract
The Risø Mesoscale PUFF model (RIMPUFF) is a Lagrangian atmospheric dispersion model that uses consecutive Gaussian puffs to simulate accidental release which is widely used in nuclear emergencies. Based on the real-time puff size and the distance between the puff and the monitoring station, RIMPUFF can calculate the dose results quickly by using the dose calculation interpolation table within its dose calculation module but the results are biased. To improve the accuracy of dose results under the requirement of rapid dose calculation, in this paper, a new interpolation table for dose calculation is proposed utilizing a 3D integral dose calculation method. And the proposed interpolation table was evaluated against the Belgian field experiment. In order to assess the quality of the optimization, the dose results of the new interpolation were compared with that of RIMPUFF, the 3D convolution method and observations from the field experiment. The comparisons indicate that the dose results of the new dose interpolation table are closer to the observed values and have better statistical metrics (0.39 for FB, 0.56 for NMSE, 0.72 for FAC2) than that of RIMPUFF (0.53 for FB, 0.76 for NMSE, 0.65 for FAC2).
The Wear and General Corrosion Properties of Low-Temperature Plasma Nitrided 17-4PH Stainless Steel With a Work-Hardened Sublayer
Authors: Han Zhongli , Lai Ping , Zhuang Wenhua , Guo Xianglong , Zhang Lefu 
Published: November 2022

Abstract
17-4PH stainless steel is used as the material of leadscrew in the reactor control rod drive mechanism (CRDM) in nuclear power plants (NPP). To improve the poor tribological properties without affecting corrosion of 17-4PH stainless steel in-service, the wear and general corrosion properties of low-temperature plasma nitrided 17-4PH stainless steel with a work-hardened sublayer were investigated. Low-temperature plasma nitriding for specimens with work-hardened sublayer was carried out at 390 °C for 1 h, 30 h. The wear test and the general corrosion test were conducted in flowing water at 280 °C and 13 MPa. The general corrosion test was carried out for 2000 h. The results show that the wear volume of the un-nitrided specimen with a work-hardened sublayer varies slightly compared to the unnitrided specimens with a polished surface (without a work-hardened sublayer), but the weight loss of general corrosion is greatly reduced (by a factor of about 3). The wear volumes of the 390 °C / 1 h treated specimen and the 390 °C / 30 h treated specimen are reduced by 3.8 and 4.4 times compared with the non-nitrided specimens (with a work-hardened sublayer), respectively, but the weight loss of general corrosion increases by 1.56 and 1.66 times, respectively.
Study on Selection Method of Shielding Nuclides in Nuclear Reactor Based on Non-Dominated Sorting
Authors: Li Rui , Liu Shichang 
Published: November 2022

Abstract
In the reactor, the main purpose of radial shielding was to protect the internal structure and staff from radiation damage. The good shielding material must have both a high elastic scattering cross-section and a high absorption cross-section to reduce the radiation level outside the reactor core and ensure the safety of workers and equipment outside the reactor. In this paper, a new type selection scheme of shielding material was proposed, which read the reaction cross-sections of all-natural stable nuclides in the ENDF/B-VII database, calculated the single group cross-sections of various reactions of all nuclides by the integral method according to the neutron energy spectrum. Then ranked the nuclides with different cross-sections as the target, to screen the nuclides with the best neutron and photon shielding effect. In this paper, a typical Monte Carlo particle transport program was used for preliminary verification, which was applied to the energy spectrum of the pressurized water reactor and sodium-cooled fast reactor. The results show that for neutron shielding, photon shielding, and neutron-photon shielding, in the thermal neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 95 %, 98.95 %, and 94.87 % respectively, and in the fast neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 72.5 %, 95.79 %, and 82.5 % respectively. The calculation time of non-dominated sorting was less than 3% of that of the Monte Carlo method. It provided an efficient method for the selection of shielding nuclides in advanced reactors and had engineering application significance.
Design Overview for Process System of Chlorine Production by Electrolyzing Saline in the Nuclear Power Plant
Authors: Wang Rui , Zeng Kehao , Xu Kechen , Yan Hongmei , Bai Wei , Li Jing 
Published: November 2022

Abstract
A nuclear power plant adopts electrolyzing saline to prepare sodium hypochlorite solution with the effective chlorine concentration no less than 8000 mg/L, which is provided to the circulating water system, essential service water system, industrial water system, potable water system, and demineralized water production system as biocide. The system name of sodium hypochlorite solution production by electrolyzing saline is the circulating water treatment system, which includes salt storage, dilute saline preparation, sodium hypochlorite solution production, sodium hypochlorite solution storage and dosing, etc. In system process design, the system automatic operation is realized by optimizing the system design and operation control scheme, and setting online monitoring instruments, automatic valves and control system to reduce the intervention of operators. In addition, the hydrogen exhaust fans are set next to the outdoor sodium hypochlorite storage tanks for mechanical hydrogen exhaust, so that the hydrogen can be quickly diluted and discharged to the atmosphere to ensure the safety of system operation. Combined with the engineering practice, this paper expounds the design overview of circulating water treatment system from the aspects of user requirements, optimization design of each process link and process equipment configuration, which can provide reference for the process design of circulating water treatment system in subsequent nuclear power plants.
Flow-Induced Vibration of Two Square Cylinders With Rounded Corners in a Tandem Arrangement
Authors: Wu Yidong , Shi Li , Wu Xinxin , Wang Xiaoxin , Xiao Qiankun 
Published: November 2022

Abstract
Based on the design of the heat transfer tube bundles with square cross section of the steam generators in high temperature reactor-pebble bed modules (HTR-PM), a free-vibration experiment is conducted to examine flow-induced vibration (FIV) characteristic of two identical rounded square cylinders with r/D = 0.12 in tandem arrangements at spacing ratio L/D = 1.5∼5.5. One of the cylinders is two-dimensional, spring mounted, and allowed to vibrate in the cross-flow direction while the other is held stationary. Considering two cases that elastic cylinder located upstream or downstream, two mass-damping ratios m*ζ are chosen to investigate the effect of spacing between two cylinder. The vibration responses and the flow structure around two cylinders are studied, using laser displacement sensor and Particle image velocimetry. It is observed that only vortex-induced vibration (VIV) occurs when the elastic cylinder is located downstream. When elastic cylinder is located upstream, the full interaction between VIV and galloping of cylinder with lower m*ζ is invariable no matter what the spacing is. The spacing has noticeable effects on the vibration behavior of cylinder with higher m*ζ and changes the interaction between VIV and galloping. The flow structure of two cylinders indicates that the upstream shear layer reattaches on the surface of the downstream cylinder as L/D &lt; 3.5, where the St drops with increasing L/D. There is gap vortices between two cylinders at L/D &gt; 3.5 where the St rises and is close to the value of single cylinder with a larger spacing.
Study on the Growth and Scintillation Properties of CLLB Crystals
Authors: Wu Kun , Zhang Leilei , Li Haijun , Huang Guangwei , Li Jiaming , Wang Libin , Zhang Siyuan , Li Zhiyuan , Wei Qinhua , Zhou Chunzhi , Wang Zungang , Liu Huilan , Zhu Hongying , Song Yushou 
Published: November 2022

Abstract
Cs2LiLaBr6:Ce (CLLB) scintillation detector has excellent energy resolution and high light output to gamma-rays, with excellent neutron/gamma-rays discrimination performance. Its applications are mainly limited to crystal growth with large size and high quality. In this paper, high performance CLLB was grown, encapsulated and tested. The vertical Bridgman method was employed to grow Cs2LiLaBr6:1%Ce crystals. By regulating the temperature, the maximum temperature gradient of the growth furnace was adjusted to the position of crystal precipitation. To avoid the non-uniform melting component area, the non-stoichiometric of raw materials increasing the excess of 100% LiBr were used. The crystals was developed successfully with 1 inch diameter, 5 cm length, and 4.32 g/mL density. Finally, the crystal was encapsulated and tested, with an energy resolution of 4.0% and relative (NaI) optical output of 1.14 to 0.662 MeV gamma-rays, and an absolute optical output of 47665 Photons/MeV to 0.511 MeV gamma-rays; The figure of merit (FOM) of the neutron/gamma-rays discrimination is 1.27 enabled by charge comparison method. The scintillation decay time of the two kinds of signals is measured respectively by the optical flux simulation method, with a fast component of about 0.19 μs and a slow component of 1.08 μs. Results show that the neutron signals has a higher fast component proportion to gamma-rays signals.
CFD Analysis of the Steam Side in a Cross-Flow X-Type Steam Superheater
Authors: Zhao Shujie , Li Weihua , Hao Wentao , Zhang Wenwen 
Published: November 2022

Abstract
A three-dimensional (3D) computational fluid dynamics (CFD) investigation of steam flow distribution of a cross-flow X-type steam superheater in a small modular reactor (SMR) has been carried out, and different flow distributors have been compared. As a non-baffle shell and tube heat exchanger, the secondary steam flow of the superheater flows across the tube bundles and is superheated in the shell side. The flow distributor is arranged between the shell side inlet and the tube bundles, devices compared are the impingement plate and orifice plate. The diameter and the geometry distribution of the holes on the orifice plate have also been researched. The software used is ANSYS FLUENT, applying standard k-epsilon turbulence flow model. Distribution of velocity and pressure drop of the shell-side steam is obtained. As the analysis of the flow-field distribution contributes to optimizing the heat exchange performance, the analysis solutions can provide helpful information for the superheater design.
Development and Preliminary Verification of Multi-Group Cross Section Processing Code MAGIC for Fast Reactor Based on Continuous Point Cross Section
Authors: Zhang Bin , Ma Xubo , Hu Kui , Zhang Teng , Huang Yuqin , Ma Xuan 
Published: November 2022

Abstract
To meet the needs of fast reactor high precision nuclear data processing, a high precision multi-group cross section processing program MAGIC based on continuous point cross section is developed, which can provide accurate multi-group cross section for fast reactor deterministic program. The MAGIC program uses 1/120 lethargy width to divide the energy interval, and a total of 2082 energy groups are divided in 0.414eV∼14.2MeV for fast reactor energy region. The main functions of the program are as follows: (1) by using the problem dependent neutron spectrum, the effect of resonance interference can be well considered. (2) The resonance self-shielding factor iteration method can be used to better consider the interaction of the resonance self-shielding in the resolved and unresolved resonance energy regions of multiple nuclides. (3) When calculating the elastic scattering matrix, the resonance phenomenon is considered in the one-dimensional elastic scattering cross section, and the prefabricated scattering transfer probability table can greatly improve the calculation efficiency of the scattering matrix under the ultra-fine group energy framework on the premise of maintaining the accuracy. In terms of benchmark verification, the effective increment coefficient was verified by fast energy spectrum critical benchmark based on the multi-group cross section generated by MAGIC. The difference between the calculated results and those of the Monte Carlo program and NJOY was small, which preliminatively verified the effectiveness of the multi-group cross section generated by MAGIC program in the calculation of deterministic method.
Eulerian Grid Approximation Algorithm of Droplet Motion in Gravity Field
Authors: Liu Zhanwei , Bo Hanliang 
Published: November 2022

Abstract
Accurate and efficient gas-liquid two-phase flow simulation is of great significance to energy, chemical industry, aerospace, environment and other fields. Although the Euler-Lagrange method can accurately simulate the motion of droplets, its computational complexity increases significantly with the number of droplets, thus it’s difficult to be applied to practical applications. Therefore, based on the explanation of the physical mechanism of the droplet movement in the steam-water separation device, combined with the Euler-Lagrange model, the Euler grid approximation model is established, which can simultaneously calculate the information of the whole droplet field and reduce the amount of computation while maintaining accuracy. In this model, the Lagrange description of droplet motion is approximated by Euler grid approximation method, and the real droplet motion is approximated as the motions of droplets along the grid nodes, so that the line between nodes is used to simulate the flow line. This paper gives the physical mechanism explanation and mathematical expression of the Euler grid approximation model to obtain the velocity, density, rotation under the gravity field, and the accuracy of the method increases with the refinement of the grid. The Euler grid approximation method can be applied to droplet motion subject to body force, as well as multi-droplet motion.