bluuebunny/crossref_metadata_2025_split huggingface dataset:
Optimization Design About Generating or Removing Vapor Space Of The Primary Loop Pressurizer
Authors: Ke Fan 
Published: November 2022

Abstract
In the second generation CPR nuclear power station, due to the limitations of design and equipment, it is not possible to generate or remove vapor space of pressurizer simultaneously with the temperature rise or fall of the primary loop coolant. But in the third generation HUALONG-1 nuclear power station, the obstacles on the equipment have been resolved and the foundation is already in place to solve the above technical problem. If we can improve the scheme appropriately, the primary loop coolant is heated or cooled while generating or removing vapor space of the pressurizer, which will be realized firstly. There are two main benefits: 1, It can reduce outage duration and improve economic efficiency. 2, It is conducive to decrease radioactive sources during the outage condition.
Construction and Lean Management of Filters Cycle Life Cycle in Nuclear Safety-Related Ventilation Systems
Authors: Du Jianxing , Deng Caiyuan 
Published: November 2022

Abstract
Nuclear safety-related ventilation systems of nuclear power plants are equipped with a large number of particle filters an d iodine filters to filter radioactive dust, aerosols and gaseous radioactive iodine. These disposable nuclear grade ventilation filters have high unit price and large annual consumption. The used filters as low and medium level radioactive waste, due to the inability to achieve reuse and effective capacity reduction treatment, resulting in a continuous shortage of low- and medium-level radioactive storage capacity in nuclear power plants. Post-processing is mainly compression temporary storage, which not only has a high cost of disposal and storage, but also has a great impact on the environment, but also has a serious waste of metal, activated carbon and other resources, which is a common problem in the industry. In recent years, in response to the national principle requirements of reducing nuclear radioactive waste and carrying out recycling, Daya Bay Nuclear Power has creatively put forward the concept of rebuilding the whole life cycle of nuclear ventilation filters on the basis of absorbing the experience of the same industry. Through the development and use of a full series of nuclear-grade reuse ventilation filters and related recycling technologies and equipment, and the lean management of the whole life process of nuclear-grade ventilation filters, the transition of nuclear-level ventilation filter from original extensive one-off life cycle management to recycling and lean life cycle management has been realized in Daya Bay Nuclear Power Base. It has embarked on a road of conservation, recycling and sustainable development.
Initial Wire-Wrapped Angle Optimization Analysis of a Liquid Lead-Bismuth Cooled Fuel Assembly Based on OpenFOAM
Authors: Wu Jie , Cai Jiejin 
Published: November 2022

Abstract
The helical wire spacer of liquid lead-bismuth cooled fast reactor (LFR) fuel assembly plays a significant role in the strengthening of the flow and heat transfer. However, most LFRs have a fixed initial angle of wire-wrapped direction in rod bundles, and the optimization analysis of their angle in the subchannel is absent. More analysis of the impacts that the wire spacer has on the liquid lead-bismuth eutectic (LBE) coolant should be obtained. In this paper, three different turbulence viscosity models and two constant turbulent Prandtl numbers were applied in the open source Computational Fluid Dynamics (CFD) platform OpenFOAM. The numerical results, which were considered with an extensive mesh sensitivity study, were validated against a series of experiment data. The simulations about key thermal hydraulic parameters such as temperature, velocity distribution, pressure drop, local and average Nusselt number were carried out based on 7-pin wire-wrapped rod bundles whose wires winding from internal, edge and corner channel. The results show that the k-ε model with Prt = 2.0 can be used to predict the flow and heat transfer characteristics of LBE. The influence of wire-wrapped starting position indicates that an optimum point exists in the internal channel. This work is useful in future safety design of fuel assemblies in the LFR.
Numerical Calculation of Ultrasonic Signal Propagation Characteristics in Steam System of Nuclear Power Plant
Authors: Zhu Jiahong , Wen Jiming , Wang Bo , Lin Yuansheng , Tian Ruifeng 
Published: November 2022

Abstract
The droplets entrained by the wet steam working medium of nuclear power plant pose a threat to the economy and safety of nuclear power plant. The humidity of steam is directly related to the working efficiency of nuclear power plant. In this paper, the ultrasonic attenuation coefficient in single-phase saturated steam is calculated. Based on the simplified ECAH model in the long wave region, the ultrasonic attenuation coefficient at different frequencies and the relationship between droplet size and humidity in the steam system of nuclear power plant are numerically simulated. For the relationship between droplet size, ultrasonic frequency, steam humidity and ultrasonic energy attenuation coefficient, the results are calculated and analyzed. Results and conclusion: through numerical simulation, the influence of hysteresis attenuation coefficient on the total ultrasonic attenuation coefficient in single-phase saturated steam medium can reach 90%. In single-phase steam medium, the fluctuation of temperature and pressure has little influence on the ultrasonic attenuation coefficient, while the deviation of ultrasonic frequency will have a large error on the ultrasonic attenuation coefficient in steam. The ultrasonic attenuation method is feasible for on-line measurement of steam humidity in nuclear power plant. Compared with the steam environment under other working conditions of nuclear power plant, the ultrasonic attenuation method is more suitable for the measurement of wet steam humidity under the conditions of humidity less than 9% and droplet size greater than 4μm.
Design of Combined Water System of Demineralized Water and Boiler Make-Up Water in Nuclear Engineering
Authors: Xu Kechen , Wang Rui , Yan Hongmei , Bai Wei 
Published: November 2022

Abstract
As an important part of the world energy system, nuclear energy plays an important role in ensuring the security of energy supply in all countries. After decades of development, the use of nuclear energy has been greatly enriched, not only limited to power supply, but also gradually played a role in the fields of Cogeneration, hydropower cogeneration and so on. The increasingly rich application scenarios of nuclear energy also put forward new requirements for the water production process of nuclear power plants. Based on the actual situation of a nuclear project, this paper introduces the water making process of demineralized water for nuclear engineering units and the water making process of boiler make-up water for heating boiler room, points out the problems existing in the original water making process scheme, analyzes the causes of unqualified steam quality of heating boiler, and puts forward practical solutions, The coupling methods of two product water production processes are discussed, and the influence of raw water quality change on the coupling process is analyzed. The coupling scheme of water production process proposed in this paper can realize the simultaneous production of product water with two kinds of water quality in one production line, effectively reduce the engineering cost of water supply facilities, reduce the floor area, ensure the safety of water supply quality, and provide new ideas and references for the multi-purpose utilization of nuclear energy.
Margin Analysis of Conventional Island Under Beyond Design Basis External Flooding Scenario for NPP
Authors: Lv Xingbing , Liu Shuangyang , Zhang Hongyang , Lin Run 
Published: November 2022

Abstract
Since the external flooding caused by the earthquake and tsunami at Fukushima Daiichi nuclear power plant in Japan, the waterproofing ability of the nuclear power plants, in operation or under construction, has attracted widespread attention from nuclear safety authorities and the public all over the world. While less attention was paid to the conventional island area. The failure of systems and components arranged in the conventional island may lead to the turbine out of the normal operation and impact the operation reliability of the unit, which would influence the generating capacity and availability of the power plant. In this paper, it is selected the combination of “design basis flood (DBF) with the rainfall of a frequency of 10-3/yr” as the analyzed scenario. According to the site characteristics, the variation of external flooding depth on the site over time was calculated under this scenario, the path of external flooding was determined, and the amount of the flooding water, the conventional island ultimately invaded, was calculated. For the availability of the power plant, the corresponding improvement measures were put forward to improve the waterproofing capacity of the plant, which can be a significant reference to the nuclear power plants in our country.
Radioactive Substance Release Risk Assessment for Fuel Mechanical Damage in Fuel Route
Authors: Wang Ming , Guo Dingqing , Wang Fei , Xu Zhixin , Zhang Bing , Wang Jinkai , Tong Lili , Cao Xuewu 
Published: November 2022

Abstract
Irradiated fuel assemblies may be damaged by dropped load or other external force during their operation and storage, resulting in a potential release of radioactive substance. In order to assess the risk of radioactive substance release risk caused by fuel mechanical damage, the methodology of probabilistic safety assessment is used in this study. It takes the initiating events analysis as the starting point, studies the various technical elements involved, and provides the corresponding quantitative assessment results and risk insights. The results of case analysis show that the dominant contribution to the fuel mechanical damage frequency comes from the drop of fuel assembly due to spent fuel pool crane failure. The risk of fuel mechanical damage can be effectively reduced by the purification of spent fuel pool water and ventilation system in the buildings. Optimizing the lifting path of spent fuel cask in hoisting pit to avoid the maximum lifting height exceeding the cask test height, which can greatly reduce the risk of spent fuel cask cask dropping.
Simulation and Experimental Verification of Boron Diffusion and Mixing in T-Tube Flow
Authors: Zhang Xinyu , Dong Bing , Yang Hui , Zhou Wentao , Wang Dezhong 
Published: November 2022

Abstract
10B has a high neutron absorption cross section, so boric acid is generally used as a neutron absorber for reactor control. During an emergency of a nuclear power system, boric acid water is injected into the reactor core by safety injection system to shut down the reactor. Boron concentration has an important impact on the operation safety of the power plant. The mixing process of boric acid and water is influenced by diffusion, convection and turbulence. A new laser-induced fluorescence method is proposed, which can accurately and quickly measure the concentration field of boric acid. This paper introduces the measurement method and the experimental study of boric acid diffusion in T-tube, and compares it with the measurement of wire mesh sensor. In addition, the CFD method is used to calculate the experimental conditions. The results obtained by the three methods were compared with each other, and the diffusion law and research of boric acid are obtained. This study can provide a certain basis for the follow-up study of boron diffusion.
Defect Detection Method for Large Size Graphite and Carbon Components of High Temperature Gas Cooled Reactor Based on Computed Tomography
Authors: Liu Renjie , Sun Yuewen , Zeng Tianchen , Liu Ximing , Sun Libin , Shi Li , Cong Peng 
Published: November 2022

Abstract
High temperature gas cooled reactor (HTGR) is a typical type of the fourth-generation nuclear power system. The main supporting structure, consisting of graphite and carbon components, play a vital role in the construction of the HTGR. The quality of the components is essential for the safety operation of HTGR since they are irreplaceable throughout the reactor lifetime. The manufacture of the components is complex, including multiple process, during which defects such as holes and crack often arise inevitably. These defects may bring serious risk to the structural safety and steady operation of the reactor. Therefore, it is of great significance to inspect and evaluate the quality of the components.
Considering the large size of the components as well as the long production cycle, traditional non-destructive testing method such as x-ray and eddy current testing are not applicable. Visual inspection and spot check are generally applied to check the surface condition, which are unable to provide the internal situation of the components. This paper proposes a helical CT based defects detection method for large size graphite and carbon components in HTGR. Graphite and carbon samples with artificial and original defect were produced, and various experiments were conducted on a multi-slice helical CT system to check the performance as well as optimize the operation parameter. The results indicates that defect larger than 2 mm in graphite components and 1 mm in carbon components can be detected and clearly visualized, which proves the feasibility of the proposed method.
Study on Improvement of Cement Curing Formula of Wet Radioactive Waste
Authors: Yan Xiaojun , Guo Xiliang , Gao Kai , Guo Xiaobin , Xi Yahui 
Published: November 2022

Abstract
In original cement curing formula, curing base material adopts the cement from the specific manufacturer. This increases the difficulty and cost of material procurement and storage in different nuclear power plants, and further leads to the waste of materials. In addition, the leaching resistance of the original formula is poor, and the cumulative leaching fraction of 137Cs is close to the values specified by the national standard. In order to make the practical work more convenient and the disposal of radioactive cement waste safer, the formulation should be improved. In this work, the cement plants around nuclear power plant were investigated, and the cement was selected as the raw material based on the requirements of the nuclear power cement solidifying formula. The coagulant calcium chloride was added to the formula to improve the delay of the concentrated solution formula. Molecular sieve was added to improve the leaching resistance of the original formula. After adjusting the proportion and composition of the formula, inclusion rate of concentrate and the waste inclusion rate for the new formula increased from 51% to 56% and from 35% to 45%, separately. Meanwhile, the compressive strength of cement waste was enhanced up to 10%. The formulation cost reduced, as calculated of the wet waste generated by two megawatt units, the solidified waste disposal cost could be saved as 2.2 million RMB per year. The ability of resisting leaching radionuclides was improved. Specifically, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the concentrated solution formula was reduced by 25.77% and 56.91%, respectively. In addition, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the resin formula was reduced by 82.80% and 82.31%, respectively.
Experimental Study of the Bistable Phenomenon in Cross Flow Over Inline Tube Bundle With P/D = 1.4
Authors: Xie Xiaoyang , Ma AnXiang , Zhao Houjian , Li Xiaowei , Wu Xinxin 
Published: November 2022

Abstract
Helical tube bundles are usually used for intermediate heat exchangers (IHXs) of High Temperature Gas-cooled Reactors (HTGRs) due to its compactness and low thermal stresses. To clarify the influences of smaller tube pitches, experiments of cross flow over tube bundle with pitch to diameter ratio (P/D) of 1.4 were carried out in a wind tunnel. The time-averaged pressure, pressure fluctuations, and velocity distributions were measured in the fully developed region of the tube bundle. The flow pattern in tube bundles will vary with the Reynolds number. The flow field tends to show an oblique flow phenomenon when Re = 10285, 18000, 44571 and 52500. The turbulent wakes over left four tube layers tend to flow left while the turbulent wakes over right four tube layers tends to flow right. When Re = 36000, the oblique flow phenomenon of the flow field diminished. When Re = 25285, two obvious quasi-stable flow patterns were observed according to the measured surface pressure and the velocity distributions measured by Particle Image Velocimetry (PIV).
PSA Application in UK HPR1000 Numerical Targets Evaluation
Authors: Xiao Lingmei , Zhang Bing , Wang Ming , Wang Jinkai 
Published: November 2022

Abstract
As an important safety assessment tool, PSA plays a very important role in identifying the weak points in Nuclear Power Plants (NPP) design and optimizing plant design. Traditional PSA risk identification is mainly carried out from two dimensions: Core Damage Frequency (CDF) and Large Release Frequency (LRF). In the Generic Design Assessment (GDA) demonstration of UK HPR1000, As Low As Reasonably Practicable (ALARP) concept and Numerical Targets (NTs) are introduced, so the application scope of PSA is expanded. Risk contribution is identified not only from CDF and LRF dimensions, but also from on-site and off-site personnel radiation risk dimensions. Based on the expanded PSA application dimension, the safety level of nuclear power plants can be further improved. This study will give the methodology of PSA application in NTs 5∼9 evaluation, and the identified NPP weak points and related design improvement suggestions of UK HPR1000.
Research on the Prevention and Control System of Jellyfish in Nuclear Power Plants Cooling Water System
Authors: Liu Xiaolin , Meng Wei , Lu Li , Li Jianwen 
Published: November 2022

Abstract
In recent years, incidents of jellyfish outbreaks invading cooling water system have occurred frequently in nuclear power plants, which directly affects the safe and stable operation of the units. In response, the prevention and control of disaster-causing jellyfish is aimed at increasing the response margin time and reducing the impact of invasion. It involves the application of multidisciplinary frontier achievements, and the design and construction of a complete prevention and control system should be based on the concept of defense in depth and systematic thinking. This paper summarizes these studies and practices, combined with the actual working conditions and needs of the nuclear power plant cooling water site, a interdisciplinary and systematic solution for the prevention and control of jellyfish in the whole chain covering the five defense lines of “prediction, monitoring, warning, interception and response” is proposed. Which is divided into the far-sea layer, offshore layer, nearshore layer, and shore layer according to the time and space elements and logical relationship to form a layered construction plan. It has the characteristics of modularization and intelligence, and can effectively improve the prevention and control of jellyfish disasters in nuclear power plants.
Implementation and Comparison of Super Homogenization Method Based on Monte Carlo and Deterministic Codes
Authors: Yuan Yuan , Liu Guoming , Huo Xiaodong 
Published: November 2022

Abstract
Monte Carlo homogenization method is increasingly used to replace conventional deterministic lattice codes to generate few-group or multi-group constants for novel nuclear systems like fast reactors and high temperature gas-cooled reactors, owing to its flexibility in geometry modeling and its wide applicability to reactors with arbitrary spectrum. In this paper, RMC, a Monte Carlo Code used for reactor physics analysis, was used to generate multi-group constants and super-homogenization (SPH) method was further used to correct cross sections for the conservation of reaction rates. In order to improve SPH calculation efficiency, SaraGR, a transport core code based on SN nodal method, was chosen to perform the iterative multi-group homogenous calculations instead of using Monte Carlo multi-group calculations. And a comparison between SPH process with SaraGR and SPH process with RMC multi-group calculations was conducted based on the very high temperature reactor critical assembly (VHTRC) benchmark. The results indicate that both SPH processes are feasible and the results agree well. The SPH process with SaraGR is much more efficient because of the inherent efficiency advantage of the deterministic method. Therefore, it is a more promising practice to generate group constants with a Monte Carlo code and perform SPH calculations with a deterministic code. Furthermore, it would be helpful to choose proper convergence criteria and introduce a proper relaxation factor into the iterative formula.
The Study of a New Efficient Monte Carlo Method for Deep-Penetration Transport
Authors: Zhang Tao , Liu Zhihong , She Ding , Zhao Jing 
Published: November 2022

Abstract
Comparing with deterministic methods, Monte Carlo method has high precision but huge time-consuming when using for shielding design. For real deep-penetration problems, a series of variance reduction methods have been proposed and applied in related software (e.g. MCMP, SERPENT) in recent decades to overcome the drawbacks of Monte Carlo method. However, these methods still have troubles, such as the selection of correction factors and function model in biasing method. The important region division method also has time and memory consuming issues in complicated models. At present, the Consistent Adjoint-Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) methods are implemented well in deeply penetrating problems. This paper presents a new efficient Monte Carlo method to solve deep-penetration problems. Contrary to traditional Monte Carlo methods, in this method, the particle trajectories that contributes to the tallies most are first determined, then the occurrence probability of the corresponding trajectory is calculated and counted. The pre-determined tracks are obtained through a serious of geometric transformations from standard tracks generated in a simple medium. The geometric transformations of tracks include rotation and stretching/shortening. Moreover, the weight correction is performed to assure the weight is unbiased. Preliminary numerical results on monolayer medium demonstrate that this method can significantly reduce calculation consumptions while retaining decent accuracies.