bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Probability Safety Assessment Method of Small Aircraft Crash Event in a Nuclear Power Plant
Authors: Wang Handing , Liang Xiaoyu
Published: November 2022

In nuclear power plants, one of the external human-caused events is aircraft crash. The method used to evaluate the risk of aircraft crash must comply with the requirements of national nuclear safety guidelines. When estimating the distance of aircraft from facility, the events for which the aircraft distance satisfying the screening guidelines may be left out of account firstly. While for those unmet, it is necessary to apply probability safety assessment to further study whether the risk of aircraft crash into nuclear power plants is within the probabilistic screening level. If the aircraft impact can not be eliminated through distance or probability screening, the next step is to identify the corresponding aircraft type, and determine the design basis of aircraft crash event as an input. For the purpose, this paper analyzes the impacts of small military airports near nuclear power plants, and researches into the probability safety assessment method for military aircraft impacting nuclear power plants, i.e., airport and non-airport operation risks. According to the study, the risk of airport operation can be ignored as the conditional probability f (x, y) representing the military aircraft falling to the site shows very low. Conservatively considering the military aircraft impact “four-factor formula parameters”, the aircraft crash impact frequency of non-airport operation is 6.23E−08/year, also the total impact frequency of military aircraft crash event, which is less than the screening level in terms of probability and meets the requirements of nuclear safety guidelines. This study aims at providing certain reference value for the establishment of design basis to protect nuclear power plants from aircraft crash.
Fuel Route Risk Assessment Framework and Pilot Plant Application
Authors: Wang Ming , Yang Chunju , Xiao Linmei , Hou Wenyu , Xu Zhixin , Zhang Bing , Wang Jinkai , Cui Lei
Published: November 2022

Damage to fuel assemblies and the release of radioactive substance may occur during fuel lifting and storage in nuclear power plant. For the purpose of quantitatively evaluate internal events, internal and external hazards induced risk of spent fuel pool and related fuel lifting facilities, to identify weaknesses of the fuel route design and provide corresponding risk insights, the methodology of fuel route probabilistic safety assessment (PSA) is systematically summarized in this study, and based on the design characteristics of a specific pressurized water reactor nuclear power plant, fuel damage frequencies and radioactive substance release frequencies in the fuel route are assessed, some possible further safety improvements for the fuel route are identified. The results of analysis case show that the fuel building ventilation system failure has the largest contribution to the overall fuel thermal damage frequency, while the spent fuel crane failure has the largest contribution to the overall fuel mechanical damage frequency. The drop of the spent fuel cask full of spent fuels in loading pit has a prominent contribution to the risk of radioactive substances release due to the mechanical damage. The safety risk in the fuel route can be effectively reduced by the diversified design of local cooling unit fans in fuel building ventilation system and type modification of the spent fuel pool crane.
Comparison of the Powderization Effect of Non-Equilibrium Plasma Oxidation and Thermochemical Oxidation Powders of Uranium Dioxide Solids for Actinide Analysis
Authors: Ma Zhuoran , Tatsuno Takaharu , Homma Yoshiya , Konashi Kenji , Suzuki Katsuya , Suzuki Tatsuya
Published: November 2022

In order to facilitate the dissolution of these insoluble nuclear debris from the Fukushima accident, it is necessary to investigate methods of pulverizing them to increase their surface area. Although solid uranium dioxide is known to become powder through volume changes caused by oxidation, thermal oxidation can lead to the volatilization of quasi-volatile radioactive materials, so it is desirable to cause reaction under the milder conditions. We therefore developed non-equilibrium plasma oxidation device to verify the powderization of uranium dioxide solids and to compare the results with thermochemical oxidation. For the results of the plasma oxidation experiment, Uranium dioxide solid (42mg) can be partly converted to powder by plasma oxidation (6.5h, O2:0.4 L/min) with a low temperature (less than 200 °C). And Partial oxidation of the powder, uranium dioxide to triuranium oxtoxide was confirmed by XRD. Small uneven bumps of 1μm or less appears on the surface of powders produced by oxidation using non-equilibrium plasma, thus increasing the surface area required for dissolution or chemical conversion to soluble compounds can be expected.
Deep-Learning Based Three-Dimensional Analysis of Bubble Flows From Light Field Images
Authors: Zhang Heng , Lyu Xue , Hang Qin , Du Yulu , Liu Bo , Liu Qun , Wang Guoyin , Liu Jinchao
Published: November 2022

An accurate measurement for the modality and motion parameters of bubbles is of great significance. In the existing bubble measurement methods, there are some problems that are desirable to be solved, such as system complexity, poor real-time performance, and the deviation due to two-dimensional measurement. To overcome these challenges, we developed a novel three-dimensional analysis method based on light field imaging diagnosis and deep learning algorithm. Different from traditional two-dimensional reconstruction, the bubble depth can be computed from light field images directly through digital refocusing technology. After calibrating, the conversion between the camera coordinate system and the real-world coordinate system is achieved through the sharpness evaluation algorithm. According to the corresponding relationship, the refocused image could be calibrated to the actual position in real world. Combined with the Multi-input Residual Convolution Neural Network, (MRCNN), the bubble depth could be computed fully automated from given images at high accuracy. Based on the above works, the three-dimensional reconstruction model for bubble flow can be established by coupling the depth and the parameters extracted through bounding boxes. The proposed method solves the problem that conventional imaging can only perform the two-dimensional measurement, which contributes to the error during the measurement process. Results show a promising performance on the three-dimensional reconstruction of bubble flow, validating the feasibility of the three-dimensional measurement method for bubbles in gas-liquid two-phase flow based on light field imaging diagnosis and deep learning algorithm.
Principal Component Analysis on Tangential Oscillation of Thermal Stratification in Pipe-Flow
Authors: Wei Dongwei , Zhou Mi , Zhang Naiqiang
Published: November 2022

Tangential oscillation of thermal stratification is a new finding in previous investigations on thermal-mixing pipe-flow at mixing tee. Due to this phenomenon, regular temperature changes can be created on the pipeline inner surface and lead to material damage caused by the degradation mechanism of thermal fatigue, which has been frequently reported as a reason for incidents on piping systems of nuclear power plants. Currently, numerical simulations have been performed to reconstruct this phenomenon by using LES (Large-Eddy Simulation) methods. The tangential oscillation can be recognized from the animation of spatial temperature distributions but is still difficult to present in writing form.

In this work, principal component analysis (PCA) has been applied to identify and analyze the tangential oscillation of thermal stratification. Time-dependent temperature data in the vertical cross-sections of stratified pipe-flow has been extracted from the previous LES calculation. The characteristics of the spatial temperature distributions has been estimated with the PCA method. The results show the angular shift of the temperature distribution and its periodical change with the physical time in the simulation. It proves that the PCA method is effective in presenting the tangential oscillation of thermal stratification and can be applied in the further investigation of this phenomenon.
Three-Dimensional Reconstruction of Single Bubble Shape and Trajectory Using Digital Image Processing and 3D Modeling Method
Authors: Zhang Heng , Liu Bo , Hang Qin , Lyu Xue , Liu Jinchao
Published: November 2022

Plate type fuel element is widely used in integral and test reactor cores with the advantages of higher power density, deeper fuel burnup and more compact structure. In the semi-closed narrow gap rectangular channel, physics-structure coupling effect under low pressure condition will lead to very complex coolant flow and heat transfer phenomena. Knowledges of the bubble movement and deformation regime are important for studying the mechanism of gas-liquid two-phase flow. In this work, a visual experimental investigation was carried out for the bubble rising characteristics and bubble three-dimensional (3D) reconstruction in a rectangular channel. The process of the bubbles motion was observed by two vertically placed high-speed cameras. By controlling the power of the oxygen pump and the aperture of the air outlet respectively, the rate of bubble formation can be controlled, which eliminated the effects of bubble-to-bubble interactions by increasing the distance between the bubbles. Based on the extracted parameters obtained by digital image processing and assumption that each cross section along the height of bubble to be an ellipse, a new method for reconstructing bubble shape was proposed, using 3D modeling method which can more closely simulates objects to connect bubble elliptic cross sections smoothly under the condition that the cross sections of bubble were not changed as much as possible. Then the bubble shape and trajectory were reconstructed. The results showed that the 3D trajectory of bubble was oscillating and the reconstructed bubble shape not only conformed to the projected contour but also satisfied the effect of surface tension which lead to smooth shape of the bubbles in any time.
Novel Syntheses Method of Grafted Chelating Resin and Its Application Study for Depleted Zinc Production
Authors: Liu Feng , He Ye , Li Jiawen , Zhan Jie , Su Xingdong , Li Xuefei , Liu Xiajie
Published: November 2022

Depleted zinc injection into the reactor coolant system (RCS) of pressurized water reactors (PWRs) has demonstrated benefits for significantly reducing radiation fields. Liquid chromatographic separation of zinc isotopes to produce depleted zinc was attempted. A novel chelating resin PS-g-GMA@IDA containing iminodiacetic acid (IDA resins) for separation of Zn2+ and zinc isotope was synthesized by γ-ray induced co-irradiation grafting of glycidyl methacrylate (GMA) onto polystyrene (PS) spherical resin, followed by ring-opening processes with iminodiacetic acid. The obtained chelating resin PS-g-GMA@IDA before and after adsorption of Zn2+ were characterized by FT-IR spectra. This showed that the adsorption of Zn2+ by chelating resin involved two processes. Firstly, zinc ions displaced the hydrogen of the carboxyl group on chelating resin, then Zn2+ chelated with oxygen atom to form a ring structure through the coordination bonds. Adsorption abilities of PS-g-GMA@IDA were evaluated by batch adsorption experiments. The effect of pH and initial concentration on adsorption capacity of PS-g-GMA@IDA were investigated. The highest adsorption capacity 130.1 mg/g for (CH3COO)2Zn was obtained at pH = 6 within 24h when C0 were 200 ppm. EDTA-(NH4)2 was selected as eluent because of its high desorption rate of zinc. A dynamic column adsorption test was carried out at 25 °C using a 1 m column packed with PS-g-GMA@IDA, and the breakthrough occurred at about 500 ml efflunent volume. Column capacity calculated from the breakthrough curve was 56.83 mg/g. Na type PS-g-GMA@IDA was utilized to carry out column chromatographic experiment for zinc isotope separation. Under the conditions of 278±1 K, migration distance 1+9 m and EDTA-(NH4)2 as eluent, the the maximum enrichment value of 66Zn/64Zn and 68Zn/64Zn in the effluent is 1.0268 and 1.0451 respectively.
Optimization Design About Generating or Removing Vapor Space Of The Primary Loop Pressurizer
Authors: Ke Fan
Published: November 2022

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Coupled OpenMC/CTF to VERA Core Physics Benchmark Problem 6
Authors: Zhang Chang , Wang Ruixiang , Guo Hui , Gu Hanyang , Xiao Yao
Published: November 2022

The multiple-physics modeling has been demonstrated to be a high-fidelity and effective method for the analysis of reactor core physics and thermal-hydraulics. OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. In this work, OpenMC and CTF are coupled for the analysis of light water reactor fuel assembly. OpenMC provides axial and radial fuel pin normalized power distribution to CTF, and CTF gives the fuel temperatures and coolant properties to the neutronics simulation of OpenMC. The windowed multipole temperature method is used in OpenMC to match the accurate temperature distribution of fuel rods and coolants obtained from CTF. In this study, the OpenMC/CTF is applied and validated using the Virtual Environment for Reactor Applications (VERA) core physics benchmark problem 6, from the Consortium for Advanced Simulation of Light Water Reactors. The problem involves a three-dimensional fuel assembly in Hot Full Power conditions. The Keff eigenvalue, pin power distribution, fuel temperatures, and coolant properties are obtained and compared with VERA’s reference results (MPACT/CTF). Our converged results showed good consistency with the reference solution, eigenvalue differences agreed within 183 pcm, axially-integrated normalized radial fission distribution difference agreed within +1.4%/−1.9%, local volume-averaged fuel pin temperatures agreed within +26.3°C, and local subchannel exit coolant temperatures agreed within +3.6°C/−0.4°C. These preliminary solutions prove that OpenMC coupled to CTF method has shown high-fidelity results in the three-dimensional fuel assembly neutronics and thermal-hydraulics coupled problems.
Application of Genetic Algorithm in Pressure Control of Pressurizer
Authors: Shan Yuzhong , Xing Kunpeng
Published: November 2022

The pressure control function of Pressurizer in the PWR nuclear power plant is one of the safety functions of the nuclear power plant. The traditional pressure control adopts the PID control method and manually adjusts the three parameters (Kp, Ki and Kd) to minimize the difference between the actual pressure value and the set value, thereby achieving the purpose of setting the three parameters (Kp, Ki and Kd) and determining the control model. The method of manually adjusting the three parameters (Kp, Ki and Kd) has low work efficiency, and generally requires several attempts to find parameters that meet the expectations, and depends on the work skills of the personnel.

In this study, Genetic Agorithm was used to convert the three parameters (Kp, Ki and Kd) into the three gene coding sequences, and the three parameters (Kp, Ki and Kd) were quickly determined through individual operation operators (selection, crossover and mutation). The research shows that the efficiency of parameter adjustment by genetic algorithm is much higher than that of manual parameter adjustment, and the optimization result is better, the control system optimized by the genetic algorithm can reach the steady state faster, has less overshoot and shorter rise time. This achievement can be considered for subsequent application in the nuclear power plants affiliated to CGNPC.
Preparation and Characterization of Radioactive Aerosol Protective Nanofiber Membranes
Authors: Li Jiawen , Chen Mingyi , Li Li , Liu Feng , Gao Yufei , Zhu Jian , Zhan Jie , Chen Jian , Zeng Yuan , Chen Rouxi , Wang Hsing-Lin
Published: November 2022

Radioactive aerosols are highly hazardous aerosols containing radionuclides. Inhalation of radioactive aerosols can lead to serious internal exposure hazards to human body. If discharged without proper treatment, it will also harm the working environment and even the ecological environment. Therefore, radioactive aerosol protection is a significant part of environmental protection and personnel protection in the nuclear field. However, majority of existing protective fabrics for radioactive aerosol filtration always meet the trade-off among filtration capacity, mechanical properties and air permeability. In this study, nanofiber layers were prepared by electrospinning technology using TPU, PVDF, PVA polymer materials and electret materials SiO2 as spinning materials. Composite membranes, prepared by coating different nanofiber layers on the PET non-woven fabrics substrate, were investigated. The results show that the 12wt% TPU nanofiber membrane has a three-dimensional spatial hierarchical structure. Its ultra-fine fiber diameter with small pore size greatly enhances the PM capture ability (PM0.3 filtration efficiency 99.99%); and the beaded spatial structure is beneficial to reduce the air resistance to 299 Pa (flow rate 95 L/min). Meanwhile, TPU nanofiber membrane has high extensibility, and it is superior to PVDF and PVA composite membranes in mechanical properties after thermal compounding. Appropriate content of SiO2 can improve filtration performance. The study shed light on developing electrospun nanofiber for radioactive aerosol protection, which can be used in the purification of ambient air in nuclear facilities, or as a high-performance fabric for radioactive aerosol protective clothing.
Development of Containment Spray Removal Model for Integrated Severe Accident Analysis Code ISAA
Authors: Li Jishen
Published: November 2022

The fission products released from the core fuel in severe accidents are mainly transported in the form of aerosols and exist in the gas phase in the containment. Once the containment fails, it will cause radioactive leakage. As an important mitigation measure for severe accidents, containment spray can effectively reduce the containment pressure and aerosol concentration, as well as the release of radioactivity to the environment. In order to analyze the thermal-hydraulic and aerosol behavior in the containment after accidents, a containment spray removal model was added to the integrated severe accident analysis code ISAA. The removal mechanism considers spray droplets washing aerosols by inertial impaction, interception collection, diffusion, thermophoresis, and diffusiophoresis. The containment spray removal model was coupled with the containment thermal-hydraulic module in ISAA to calculate the rate constant for spray removal. Benchmark experiments were selected for assessment of the improved code. The comparison shows that the results of the thermal-hydraulic response and aerosol mass distribution simulated by ISAA are consistent with the experimental data trends. The improved code can accurately simulate the thermal-hydraulic response and aerosol mass distribution during aerosol removal by spray droplets. Implementation of the containment spray removal model in ISAA satisfies the analysis of aerosol behavior in the containment.
The Research on the Interception Engineering Layout of Active Intercepting and Guiding in Water Intake Open Channel of Nuclear Power Plant
Authors: Yang Jia , Cao Ranran , Bai Wei
Published: November 2022

At present, most intercepting facilities in water intake open channel of nuclear power plant are passive. Moreover, the correlation between the layout of intercepting facilities and hydraulic conditions of water intake open channel is not considered. Hence, the layout of intercepting facilities is unreasonable. The cooling water intake system of nuclear power plants faces risks. In this paper, The flow field characteristics of water intake open channel under typical sea hydrologic conditions were studied. Then, the migration path and aggregation rule of floating or suspended objects are simulated along with the flow. The blockages-removal effect of floats and marine organisms are studied after different intercepting engineering measures are carried out in water intake open channel. The technical route combining mathematical model and physical model test is adopted. Finally, the principle and scheme of the interception engineering layout of active intercepting and guiding are proposed. The optimized interception engineering layout can effectively guide suspended or floating blockages to the outside of water intake, or guide blockages in water intake open channel to the area without disaster risks, where the blockages are also easy to be salvaged. It can significantly reduce the large quantity of blockages into the cooling water intake system, which will also reduce the blockage loads for all intercepting facilities. Thereby, the defense capability of nuclear power plant cold source system against potential water intake blockages is greatly improved.
A Method to Evaluate Defect Size of Failed Fuel Rods in CGN 1000MWe PWRs
Authors: Fu Pengtao , Li Zhijun
Published: November 2022

Fuel cladding is the first barrier to confine the release of fission products to the environment during normal operation. Once the fuel rods defect, the fission product accumulated in the pellet-cladding gap of failed fuel rods will release into the primary loop and lead to high activity levels in a pressurized water reactor. It is necessary to determine the status of fuel failure to avoid its degradation in the operation.

This paper introduces the mechanism of the generation and release of fission products in fuel rods and the primary loops in a pressurized water reactor, and provides the theoretical method of the regression between release-to-birth R/B and decay constant λ of radio-iodine. It has been verified by one typical case of fuel failure in CGN 1000 MWe PWR, but the post-irradiation examinations in hot cell reveal that six defects exist in one failed rod and the actinides have an obvious release from the failed rods due to secondary hydriding. The actual status of defect size of failed rods is much more complicated than expected in most analysis codes and modelings and it is suggested to introduce some parameters, e.g. the equivalent defect size, to describe the status of fuel failure in future research. The related operation experience in CGN has also been presented to help to better understand the status of fuel failure in PWR units.
Partially Averaged Navier Stokes Simulations of Turbulent Thermal Mixing in T-Junction
Authors: Bilal Ashhar , Gao Puzhen
Published: November 2022

High Cycle Thermal Fatigue is a major degradation mechanism associated with the nuclear power plant piping where hot and cold streams of fluids mix together with a phenomenon called turbulent thermal mixing. It induces significant thermal fluctuations on the piping wall called thermal stripping. T-junction has been known as a typical component susceptible to high cycle thermal fatigue. Accurate prediction of turbulent flow and thermal fields at the wall is necessary to predict the thermal fatigue degradation in such piping systems. Computational Fluid Dynamics (CFD) with advanced turbulent modelling is required to predict the flow and temperature fluctuations at the wall. Wall resolved LES is typically known to give good results for velocity and temperature fluctuations for such a flow phenomenon. However, LES is computationally expensive, and the cost of computation increases with the higher Reynolds number flows. Partially Averaged Navier Stokes (PANS) is a Hybrid turbulence Model. It is a bridging method from Reynolds Averaged Navier-Stokes (RANS) to Direct Numerical Simulation (DNS). PANS is usually controlled by two resolution control parameters, fk (unresolved to total kinetic energy) and fε (unresolved to total dissipation), fk = 0 being the DNS and resolving all scales of turbulence, and fk = 1 is RANS simulation. In this study PANS simulations for the Vattenfall T-junction Benchmark case are performed. fk is considered constant throughout the domain. Average and Root mean square temperature and Velocity profiles are compared with experimental data. It is concluded that PANS model can simulate turbulent thermal mixing phenomenon in a T-junction.