bluuebunny/crossref_metadata_2025 huggingface dataset:
Research on Sensing Value Reconstruction Method of Small Modular Reactor Based on Improved Short-Term and Long-Term Memory Neural Network
Authors: Cheng Shouyu , Zhu Hao , Zhang Bowen , Sun Jilin
Published: November 2022

Small modular reactor (SMR) is a complex, time-varying and strongly coupled nonlinear system, which adopts integrated design with short pipeline and small coolant inventory. Sensor failure of SMR under frequent load-following conditions threatens the safety of reactor. It is an effective way to realize sensor fault tolerance by abnormal sensing value reconstruction with machine learning method. In this paper, a sensing value reconstruction model combining hybrid dilated convolution neural network (CNN) and long short-term memory (LSTM) neural network is proposed. The hybrid dilated CNN is used to extract the feature information in the data set as the input of LSTM. LSTM is trained through training set to obtain the reconstructed sensor parameter time series. Besides, the traditional LSTM model is trained as a comparison model to verify the advantages of the reconstruction model. The training set is obtained from the SMR simulation model established by RELAP5 code. Multiple sensing values are input to each other’s reconstruction model. In order to avoid the one abnormal sensing value affecting the reconstruction behavior of other sensing value reconstruction models, the abnormal sensed value is replaced by its reconstructed value as the input of other models. The test set completely different from the training set is used to verify the behavior of the reconstruction model. The results show that the reconstruction model proposed in this paper not only fit the real data well, which means good reconstruction behavior on sensing value reconstruction, but also have better reconstruction effect compared with the traditional LSTM model.
Thermohydraulic Characteristics of Supercritical CO2 Flowing in Printed Circuit Heat Exchanger With Large Eddy Simulation
Authors: Jin Feng , Chen Deqi , Hu Lian
Published: November 2022

Printed circuit heat exchanger (PCHE) is a promising candidate of supercritical CO2 (S-CO2) Brayton cycle in the IV nuclear reactor with high efficiency, compactness and high resistance of temperature and pressure. In this paper, model of zigzag PCHE (D = 2.0 mm, Lp = 7.24 mm and θ = 100°) was simulated under the precooler condition near and away pseudo-critical region with S-CO2 and water as hot and cold fluid respectively. Method of large eddy simulation (LES) was applied to investigate the local performance of S-CO2 affected by both of complicated structure and drastic properties variation. The results showed that in the local position of a pitch, temperature and velocity on cross-section presented strong nonuniformity due to the disturbance of corner and symmetry on the side of corner, which showed heat transfer enhancement or weakness on different locations. Meanwhile, the turbulent structure of S-CO2 showed a periodic characteristic among pitches without influence by the drastic change of thermo-physical properties. Based on the results, two flow regions, convection region and mixed region were defined by dividing the local pitches. Combined with the vortex structure analysis using Q criterion, it was found that the mixed region presented more intensive vortex distribution and thus showed a higher heat transfer performance and flow resistance while convection region showed the opposite trend. Therefore, a fluctuation changing curve along the flow direction were observed on heat transfer efficiency and friction factor due to different characteristics of two regions.
Study on the Response of Prestressed Reinforced Concrete Slabs With High Reinforcement Ratio Under Low Speed Impact
Authors: Song Mengyan , Jiang Di , Li Changkai , Zhang Chaoqi , Zhi Xudong
Published: November 2022

Since the 9.11 incident, the impact of large commercial airliners on nuclear power plants and other important national facilities has attracted worldwide attention. The safety of flying objects hitting nuclear containment has become one of the important research topics in the field of nuclear power generation. In this paper, the impact resistance of non-prestressed and prestressed concrete slabs is studied through detailed experiments and numerical studies. By non-bonding rear sheeting, prestress of about 10MPa is applied to concrete slabs measuring 1400mm × 600mm, which are then hit by hemispherical steel hammers (590kg) at heights of 2m and 4m above the center of the span. The impact force, displacement, and the reinforced concrete slab and the prestressed concrete slab are compared. In LS-DYNA finite element software, the CSCM concrete structure model is used to study the numerical values of the experiments carried out. It has been found that with the increase of impact energy, the deformation produced by reinforced concrete slabs increases and is not recoverable. On the other hand, after applying presession, the impact force of prestressed concrete slabs was increased by 53.8% and displacement was reduced by 45% due to the increase of the stiffness of the plate. The maximum reproduction deviation of the impact force and displacement values is 17% and 14%, respectively. The research results will be applied to the structural dynamic impact protection design of nuclear power plant containment.
Interpretation and Analysis of Design Idea of the Spent Fuel Transport Ship
Authors: Zhong Gui , Hui Zhang , Wei Wang , Yueyan Yuan , Chao Liu , Xiaoqing Li
Published: November 2022

With the rapid growth of China’s nuclear power scale, in order to meet the future demand for spent fuel transport, relying on a single road transport mode is far from meeting the transport demand. It is necessary to build a spent fuel multimodal transport system to carry out large-scale spent fuel transport. As a special means of transport for sea transport in spent fuel multimodal transport, China’s first spent fuel transport ship is completely independently designed by China, has passed the empty ship trial voyage, and has also achieved good practice in actual transport. Based on the design requirements of domestic and international laws and regulations for the spent fuel transport ship, and combined with the actual situation of the wharf of domestic nuclear power plant, this paper interprets and analyzes the design ideas of the ship type size, loading layout scheme and safety system of the first spent fuel transport ship in China, in order to provide technical reference for the design and improvement of the spent fuel transport ship in the future.
New Opening Strategy of Filtered Containment Venting System
Authors: Zheng Hua , Wei Shuhong
Published: November 2022

Filtered containment venting system (FCVS) is used in some nuclear power plants to avoid containment overpressure failure and uncontrolled radioactive release in severe accidents. Currently the opening strategy is mainly based on the containment pressure, such as the design pressure. The containment has a stronger capability. There are various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment. There are various natural or engineered deposition mechanisms of aerosol particles in the containment. The qualification pressure of FCVS can be increased. If the FCVS is opened later, the total radioactive release into the environment can be decreased, even though the leakage may be increased. A new opening strategy of FCVS is proposed, which is not just based on the containment pressure, but considers the containment capability, various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment, various natural or engineered deposition mechanisms of aerosol particles in the containment. This strategy uses the radioactive release into the environment and associated radiological consequence as the optimization objective, considering both radioactive release through potential micro channels before venting and radioactive release after venting, and can reduce the radioactive released into the environment as far as possible while maintaining the integrity of the containment.
Research on the PID Parameter Self-Tuning of the Pressurizer Control System of Nuclear Power Plant
Authors: Xing Kunpeng , Wang Hongrui , Zou Tao , Yang Zhijia , Wang Jingyang , Zhang Pengbin
Published: November 2022

The pressurizer is an essential part of the safe operation of the nuclear power plant. It contains both the steam phase and the liquid phase, which requires simultaneous control of the pressure and water level, and the control system is strongly coupled. When using the PID control method, it is difficult to tune the parameters, and the difficulty is exacerbated when the set value fluctuates. This paper aims at the multi-input and multi-output pressurizer control system, based on the particle swarm optimization to realize the PID controller parameters self-tuning, adopting the time and absolute error integral as the fitness function to evaluate the merits of the PID parameters, and constantly iterating to update and obtain the PID parameters that better satisfy the current working conditions. For coupled pressurizer control systems, the PID parameter tuning method using particle swarm optimization algorithms can give a group of PID controller parameters that make the system operation stable despite the frequent fluctuation of the set value, and compared with the traditional PID controller parameter tuning method, this method can realize the PID parameter self-tuning, eliminating the trouble of manual tuning.
Thermal-Hydraulic Analysis of HL-2M Cooling Water System for Toroidal Field Coils
Authors: Wang Jiaqi , Xie Yanfeng , Liu Xiaolong , Meng Jianpeng , Jia Ruibao , Li Qiang , Chen Xin
Published: November 2022

HL-2M is a medium-sized copper-conductor tokamak from the Southwestern Institute of Physics (SWIP). HL-2M cooling water system is composed of Tokamak Cooling Water System (TCWS), Component Cooling Water System (CCWS), and the Heat Rejection System (HRS). The tokamak cooling water system, the primary cooling system of the tokamak machine, is further divided into TCWS-1, TCWS-2, TCWS-3, TCWS-4 depending on the water chemistry needs of clients and pressure requirements. As an important subsystem, the function of the TF coils cooling water system (TCWS-1) is to remove heat load from Toroidal Field coils. HL-2M copper TF coil is made up of 20 coil bundles, each consisting of seven turns for 140 turns, each of turns with six cooling channels for 840 cooling channels. A thermal-hydraulic model of TCWS-1 magnet coil client was established, by COMSOL Multiphysics, to conduct the transient state thermal-hydraulic analysis. The TCWS-1 cooling loop was developed, by using AFT Fathom code, to conduct the steady-state thermal-hydraulic analysis of the system. In this thermal-hydraulic analysis model, by multi-physical coupling calculations on the client and applying the results to the TCWS-1 loop calculation, more realistic results can be obtained. This paper presents the results of this thermal-hydraulic analysis which was composed of the main thermal-hydraulic parameters for TCWS-1 loop and client, including flow rate, velocities, pressure drops, and outlet temperatures. The results obtained in this study can provide support and reference to the engineering design of TCWS-1.
Study on Separation of Palladium From High Level Liquid Waste by Potassium Manganese Ferrocyanide
Authors: Qi Tianming , Sang Hongji , Mao Cong , Wen Yueying , Hu Jipu , Wu Yan
Published: November 2022

Glass solidification is a favorable treatment method for high level liquid waste (HLLW) from spent fuel reprocessing in industrial application, and the presence of palladium (Pd) in HLLW may seriously affect the glass solidification process. Ferrocyanide with a perovskite-like face-centered cubic structure has strong adsorption affinity towards Pd. In this study, silica-based composite KMnFC/SiO2 was prepared by pore crystallization of potassium manganese ferrocyanide (KMnFC) into porous SiO2. A series of characterization of the synthesized adsorbent were carried out such as XRD, TG-DTA, and SEM-EDS. It was verified that the adsorbent was successfully prepared and it has good Microscopic structure along with element distribution.

A series of adsorption experiments were carried out. The adsorption of Pd is not easily affected by the concentration of nitric acid, which shows good acid resistance. Through the adsorption isotherm and adsorption kinetic curve, it is proved that the adsorption process of Pd by KMnFC/SiO2 is chemical monolayer adsorption. The adsorbent has strong adsorption selectivity for Pd in multi-ionic solution. Under the condition of complete adsorption of Pd, the adsorption rate of the adsorbent for other elements except Ru is less than 20%.
Improvements on Fuel Failures Diagnosis Method for Pressurized Water Reactor Nuclear Power Plant Based on Radioactive Activities Inside the Waste Gas Treatment System
Authors: Lyu Weifeng , Guo Runchun , Ran Wenwang , Gao Yaoyi , Xiong Jun
Published: November 2022

The cladding of fuel elements in the core of a pressurized water reactor nuclear power plant (PWR) is the first barrier to the radioactive material in the reactor. Once the fuel cladding fails, the radioactive fission products or even fissile material in the fuel elements will be released into the primary coolant, thereby challenging the safe operation of the plant. Fuel cladding failures diagnosis is indispensable for PWR as fuel failure is not inevitable and the impact on nuclear safety is significant. Based on the continuous research on fission product generation and release mechanisms for decades, several fuel failure diagnosis methods for PWR have been established and specific nuclides have been selected as indicators of fuel failures. Recent research shows Kr-85 is a good indicator for the diagnosis of the number of failed fuel rods for its insensitivity to the fuel rod defect size and failed fuel burn-up, but it is hard to measure the activity of Kr-85 in the primary coolant due to the interference of other nuclides. However, the activity of Kr-85 inside the waste gas treatment system can be measured where the activity of other nuclides has been significantly reduced by decay. Hence, the fuel failure diagnosis method is improved based on Kr-85 activities inside the waste gas treatment system. The operating data of Chinese in-service nuclear power plants are used to validate the improved fuel failure diagnosis method, and the validation results show that the improved diagnosis method of fuel failures can accurately diagnose the fuel failure and has wider applicability.
Research and Development of Default Value Analyses and Management Platform for Nuclear Power Plant Control System
Authors: Zhao Hongbin , Liu Zhiyong , Liu Hemin , Liu Weifang
Published: November 2022

The research and development of default value development and management platform of nuclear power plant control system can realize the functions of default value analysis, value selection, query and data management. The software standardizes the process and judgment factors of default value analysis process, makes qualitative and quantitative judgment on value selection factors, and combines the value selection algorithm embedded in the software, The auxiliary designer can automatically get the default value. The use of software can effectively improve the rationality and standardization of default value analysis. The software platform solves the problems of different default value analysis standards and chaotic data management in nuclear power plants. As the first special software for signal default value analysis and management of nuclear power control system in China, it has improved the intelligent level of the third generation nuclear power HPR1000.
Research on Measurement Method of a Device for Measuring the Perpendicularity of Tubesheet Hole
Authors: Xu Xianghua , Hu Guihong
Published: November 2022

The tubesheet is the key part of the steam generator, and the quality of the tubesheet hole is the key characteristic of the steam generator. The inspection of tubesheet holes includes defects on the surface of the tubesheet holes and the geometric dimensions of the tubesheet holes. The geometric dimensions of the tubesheet hole mainly include hole diameter, perpendicularity, hole bridge, position and roughness, etc. The perpendicularity of tubesheet hole is an important characteristic, which directly affects the size of tubesheet secondary hole bridge and heat exchange tube installation. This paper describes a device for measuring the perpendicularity of tubesheet hole, discusses the principle and measurement method of the tubesheet hole perpendicularity measuring device, and analyzes the error and proposes measures to improve the measurement accuracy of the device. Analysis of data processing provides a basis for the measurement device and measurement method to meet the requirements of specifications. The results show that the designed measuring device to measure perpendicularity has the advantages of simple operation, simple measurement and high accuracy of measurement results. It is beneficial to ensure the quality of heat exchange tube installation of the steam generator, which can speed up the manufacturing progress of the steam generator.
Practical Issue Analysis of Miniature Circuit Breaker Implementation
Authors: Xingwei Fei , Kunpeng Tian , Yu Zhao , Ping Li
Published: November 2022

Miniature circuit breakers are widely used in engineering applications. This paper analysis the operation principle of overload protection and short circuit protection from its structure. Based on the actual situation in unit 3 and 4 of Qinshan Nuclear Power Plant, the basic reasons for several typical problems which have occurred on the miniature circuit breaker are analyzed and the treatment scheme is given respectively. Aiming at the defect that the breaker incorrect triped in main transformer cooling control-box in the high temperature environment, combined with the temperature coefficient correction and the parallel arrangement of the drop capacity effect, the miniature circuit breaker is rechosen to avoid the defect. In view of the defect that the upstream breaker of the PAMS cannot be switched on during the surge current, by giving the analysis of the match between the electromagnetic release characteristics of the miniature circuit breaker and the impact current of the downstream parallel load closing, the scheme is put forward to increase downstream parallel load breakers. According to the selectivity of the miniature circuit breaker, the unreasonable problem of the last stage distribution in the protection configuration is found out in view of the problem of RPR protecting the internal short circuit of the cabinet, and according to the requirements of the national standard, the selective test verification of the related model miniature circuit breaker is carried out. The reasonable part has been changed and carried out on the spot. After the actual operation of the unit, three typical problems occurred on the miniature circuit breakers have been effectively solved.
Design Improvement of the Discharge Proposal Preventing SG Overfill After SGTR in HPR1000 PWR
Authors: Li Qinghua , Fang Jun , Li Yafei , Zhao Bin , Wang Guangfei
Published: November 2022

Base on the result of full scope accident analysis for HPR1000,which is a third generation localized PWR with combination of initiative and passivity, the secondary side pressure of the damaged SG should be controlled by the steam generator blowdown system (TTB) during the retreat process of SGTR accident. In order to response the safety criteria of DBA function, the proposal of post accident discharge pipeline of TTB system should be improved. The part of secondary side fluid are discharged through the pipeline after SGTR, to realize the decrease of water level of damaged SG and preventing SG overfill. The paper suggests a proposal of post accident discharge pipeline for TTB as a mitigation measure of SGTR accident. The discharge flowrate with two phases flow condition in the discharge pipeline must be verified by theoretical calculation to evaluate the mitigation capability of the proposal. Now the performance test has been finished in the first unit of HPR1000, The achievement indicate the discharge capability of the proposal conform to the relative safety criteria for the SGTR accident.
General Safety Assessment of Seismically Induced Hazards for NPPs
Authors: Yang Guanghua , Yang Cuilian
Published: November 2022

Design basis earthquake may cause various internal hazards in nuclear power plants, such as fire, explosion, flooding, missiles, high-energy pipelines break, etc. In addition, earthquakes may also cause non-seismic items to fall or collapse, thus affecting safety-related systems or equipment. As the subsequent situation induced by earthquake is quite complex, no practicable methods have been presented yet to be used for the safety evaluation after earthquake. This paper introduces a feasible safety assessment concept for the above-mentioned hazards caused by the design basis earthquake based on the engineering practice experience.

This general assessment concept based on the deterministic approach could be used for guidance of safety evaluation during or after earthquake for nuclear power plants.
Research on Development of On-Line Monitoring of Fuel Damage in Nuclear Power Plant
Authors: Yiduo Tian , Wenjun Hu
Published: November 2022

The occurrence of fuel damage has serious impact on safety and economic efficiency of nuclear power plants. The on-line monitoring and diagnosis system of fuel damage in nuclear power plant can timely determine whether the fuel element cladding is damaged or not, and diagnose the relevant information such as the size of the rupture, the position of the damaged component and the location of the damage.

The paper investigates some fuel damage monitoring systems and analyses their basic structure and process. Based on the examples, the paper classifies the process into radioactive signal detection, decomposition and failure diagnosis part and discusses the theoretical and experimental basis. Different methods of fuel damage diagnosis along with factors resulting in errors and applicable conditions are also discussed.

Through the investigation, some results can be concluded that the reason that the online detection system of fuel damage is not widely used is mainly due to precision limitation. Future efforts should be made on decomposition and failure diagnosis part in order to reach a fast response and high accuracy.
The Influence of Impurity Water for CHF on the Downward Facing Boiling Surface
Authors: Lin Bo , Yun Xiangyu , Xu Junying , Wei Dongshan , Zhang Huiyong
Published: November 2022

In-Vessel Retention (IVR) strategy is one of the most important severe accident management strategies. The Critical Heat Flux (CHF) is an important key element in the IVR research. How to increase the CHF get more and more attention in the CHF research. This paper aims at the influence of the coolant with different impurity. Those impurity can fall into the coolant when the severe accident happens. In this experiment, the deionized water was used as the coolant. The experiment investigated the different impurity with boronic acid that has different concentration and the boronic acid water with different impurity, such as concrete, fiberglass and coating. The concrete can be fallen from the construction, the fiberglass can be fallen from the insulation cotton which cover the pipe and device, and the coating can be fallen from the devices and other constructions. When the accident happens, those impurity can fall from the construction, then fall into the coolant and influence the CHF of the outside surface of lower head. From the result of the experiment, the CHF with the boron acid water is higher than that only deionized water. The CHF with the impurity water depend on the concentration of the impurity. Lower concentration will increase the CHF when the higher concentration will decrease the CHF. Different impurity has different influence on CHF, we control the component of the coolant to improve the CHF.
Research on Immersive Simulation of Key Equipment Process Flow in Nuclear Power Plant Based on VR Technology
Authors: Zhang Yi , Zhou Zhi-Gang , Wang Hao , Yang Zheng-Hui
Published: November 2022

Nowadays, 3D modeling and virtual reality technology has become a new method to enhance the digitization level in nuclear power engineering field. Reactor pressure vessel head handling, refueling and main equipment installation are the important process flows in nuclear power plant. Traditional measures without digitization and visualization, which need much more experience and technical skills to conduct the staffs on-site construction work. In that case, there are more risks when they do some complicated operations especially the new ones they never know before. This paper presents a technical method to design an immersive virtual environment, which contains the whole 3D model data of the nuclear island buildings in nuclear power plant to simulate the key processing flows. Meanwhile, the paper develops an interactive VR software system to help training workers about their on-site works in a virtual way before they solve on-site problems. On the other hand, another application of this paper’s work is design verification of the process flows themselves, so that the simulation results can tell designer what and where is the optimizing goal. The achievements of this paper have already been applied in the actual HPR1000 nuclear power project of CGN, and bring considerable efficiency return in construction field.
A Comparative Approach of a Seismic Response Analysis Based on Housner Model and Added Mass Model for a Nuclear Water Storage Tank
Authors: Yang Chengzhuo , Metais Thomas , Zheng Zihao , Dai Yang , Min Jiesheng
Published: November 2022

Water storage tanks are widely used in the nuclear industry for safety purposes, chemical water control and primary volume water management. Under seismic load, the sloshing of water inside the storage tank can cause a strong dynamic response, which may also change some mechanical characteristics of the tank, for example the center of mass. The water sloshing analysis is therefore an important part of the modal and seismic analysis of water storage tank. However, the water sloshing effect contains a complicated fluid-structure interaction but structural analysts cannot afford the time to spend on the calculation. It is important to build up a simplified equivalent model for the simulation.

EDF has been developing since 1989 its own in-house FEA code baptized code_aster, which is included in the SALOME-MECA mechanical package. In this paper two simplified fluid-structure interaction models are performed using SALOME-MECA. Firstly, an optimized Housner model is derived with formula and implemented with 3D finite model. Additionally, another 3D finite element model using added mass model assembled in code_aster, which can simulate more precisely the fluid-structure interaction of the water storage tank.

Based on the theory of these two models, their natural modes results and seismic response results are carefully compared and analyzed. The results indicate Housner model can provide a precise value of the water sloshing frequency and its displacement value is not accurate enough due to the impulse mass and convection mass simplification. However, the added mass model in SALOME-MECA can provide a more accurate and reasonable result on vibration mode calculation, which may provide an interesting alternative method for seismic verification for the nuclear refueling water tank.
Dynamic Wetting Behavior During Droplet Impacting on a Superheated Surface
Authors: Chen Mingjing , Chen Deqi , Liu HanZhou
Published: November 2022

During the reflooding process after a loss-of-coolant-accident (LOCA) scenario, the highly superheated fuel surface of the reactor core will undergo a quenching process. The temperature of the fuel element surface drops sharply during the quenching process, because large droplets can pass through the vapor film to re-wet the wall surface, and the boiling mechanism is transformed from film boiling to nucleate boiling, the boiling heat transfer perfomance is greatly improved. But the transition of the boiling mechanism is closely related to the initial velocity of the droplet, the size of the droplet, and the wall temperature. At the same time, after the droplets hit the wall surface, dynamic wetting behaviors such as deposition, vibration, rebound, or splashing will occur. Unfortunatly, the mechanism of droplet wetting behavior on boiling heat transfer is not clear, so it is necessary to carry out a fundamental research. In this study, the droplet size, initial droplet velocity, substrate’s temperature and surface tilt angle were precisely controlled. The images were captured by a high-speed camera to obtain a dynamic wetting image of the droplets impacting the wall from room temperature to 400°C. Finally, based on the image analysis method, a simple analysis of the three-phase contact line movement and the droplet dynamic wetting behavoir is performed.
Phase Distribution Characteristics of Bubbly to Slug Flow Transition Region in a 10mm ID Vertical Round Pipe
Authors: Xu Wangtao , He Qingche , Huang Tao , Ding Shuhua , Yan Meiyue , Zhang Luteng , Pan Liangming
Published: November 2022

As a common heat transfer structure, the vertical round pipe is widely used in nuclear energy systems. The phase distribution characteristics of two-phase flow in a vertical round pipe is an important issue. For two-phase flow, combined with the local typical flow phenomenon, the analysis of the phase distribution characteristics such as the void fraction and the interfacial area concentration is helpful to research the two-fluid model and interfacial area transport equation. Based on the completed flow pattern experiment, an air-water two-phase flow characteristic experiment is carried out for the visualized experimental section of 10 mm ID vertical round pipe. In addition, the 5 × 7 total 35 measuring points of bubbly to slug flow transition region are selected, including 5 superficial liquid velocities (0.5, 1.0, 1.5, 2.0, 2.5 m/s) and 7 superficial gas velocities (0.08, 0.11, 0.13, 0.17, 0.21, 0.28, 0.35 m/s). High-speed camera is used to capture the flow patterns of bubbly to slug flow transition region. Four-sensor conductivity probe is used to measure the interfacial parameters such as void fraction and interfacial area concentration. Thus, the phase distribution characteristics of bubbly to slug flow transition region in the round pipe is analyzed. The obtained two-phase flow parameter data can provide data support for the development of two-fluid model and interfacial area transport equation.
Investigation of Design Models for Printed Circuit Heat Exchangers With Straight Semicircular Channels
Authors: Mao Junjun , Zhao Gang , Yang Xiaoyong
Published: November 2022

Printed circuit heat exchangers (PCHEs) are a promising technology for the closed Brayton cycle of nuclear energy due to their high compactness, capability to withstand high pressure and large temperature difference, and high heat transfer efficiency. The design calculation of PCHEs has a significant impact on the cycle performance. This paper proposes a theoretical model for predicting the thermal-hydraulic performance of semicircular straight-channel PCHEs. The irregular fins are equivalent to rectangular fins of equal height, making the model more simplified and accurate. The model is verified using experimental data from the literature. Based on this model, the heat transfer characteristics of composite PCHEs, where crossflow exists at the inlet and outlet, are investigated. The results indicate that the heat transfer performance of PCHEs can be more accurately predicted by using the length of the overlapping region of hot and cold channels as the reference length. Furthermore, the effects of geometric parameters on the heat transfer performance of semicircular straight-channel PCHEs are studied. The results show that the heat transfer performance of the semicircular straight-channel PCHEs decreases with the increase of the channel radius and the vertical interval, and increases with the increase of the horizontal interval. And the horizontal and vertical interval affect heat transfer performance slightly. This study provides a reference for the design and application of PCHEs in the helium closed Brayton cycle in the future.
Safety Analysis for Transportation of Activated Secondary Neutron Source Assembly
Authors: Cao Nanfeng , Yang Shouhai , Xie Enfei , Jing Shunping
Published: November 2022

With the improvement of economic requirements of nuclear power operation, China General Nuclear Corporation (CGN) will gradually replace the more expensive primary neutron source assembly with the activated secondary neutron source assembly in the first cycle for reactor start-up. After activation, the secondary neutron source assembly needs to be adapted to the simulated fuel assemblies by underwater operation, and then integrated into a dedicated secondary neutron source assembly transport cask and transported to the reactor core. Therefore, the safety analysis for transportation of activated secondary neutron source assembly is important. Based on the NAC storage transport cask, the Monte Carlo program and CFX software are used for safety analysis mainly including structural evaluation, thermal evaluation and shielding evaluation. The result of structural evaluation shows that the containment is not breached under any of the normal conditions of transport or the hypothetical accident conditions. The result of thermal evaluation shows that the decay heat released by the activated secondary neutron source assembly can be derived effectively by the transport cask and the temperature of each component cannot exceed its corresponding limit. The shielding calculation results also meet the shielding acceptance criteria. Thus, this work verifies that the transportation of activated secondary neutron source assembly meets the design requirements in the main aspects of nuclear safety and is feasible.
Cause Analysis and Treatment Measures for Corrosion of Warm-Up Pipe of Hualong no.1 Main Feed Water Pump System
Authors: Wu Zhiming , Ling Yixin , Liu Xing , Fei Dongdong , Hou Xiaoyu , Wen Xue
Published: November 2022

The Main Feedwater System of Fuqing nuclear power plant is composed of three electric pump sets arranged in parallel with 50% capacity to provide the required Feedwater for the secondary side of steam generator. Under normal conditions, two pumps are in operation, one is standby, and three pumps can be switched at will. When one of the two running electric feed water pump sets trips, the electric feed water pump set in standby state will start quickly. If the equipment has problems, it will seriously affect the safe and reliable operation of the unit.

After the leakage defect of the warm-up pump pipe downstream of the warm-up pump stop valve in train C of the main feed water pump system of Hualong unit was found, through the disassembly inspection, it was found that there were many surface corrosion pits in the downstream pipeline of the throttle orifice plate. In this paper, through the physical and chemical analysis of the pipeline composition, organization, non-metallic inclusions, surface quality, macro and micro morphology, etc, At the same time, combined with the operating conditions of the pipeline, analyze the causes of the defects, judge the defects caused by the medium vaporization caused by the excessive throttling of the throttle orifice, and give effective treatment measures. Through these relevant measures, the corrosion rate of the warm-up pump pipe is reduced and the operation reliability of the important equipment of the unit is improved.
Engineering Application and Research Progress of Low-Level Radioactive Waste Incineration Technology in China
Authors: Haoran Chu , Wei Xu , Bowen Zheng , Jiasheng Ruan
Published: November 2022

China Institute for Radiation Protection had independently developed a multi-purpose radioactive waste pyrolysis incineration technology and built 3 incineration facilities in China, which were mainly used to treat solid waste and waste oil from nuclear facilities such as NPPs. The composition of solid waste included paper, cloth, plastic, rubber, etc. In order to solve the problems in the early operation of incineration facilities, targeted improvements had been made in the aspects of equipment anti-corrosion capability, secondary waste generation and system safety. The improved facilities had treated a large amount of low-level waste and operated for more than 15 years. The stability and reliability of the incineration system were verified, and the advancement of pyrolysis incineration technology was proved.

Considering the current situation that the proportion of plastics in low-level waste is increasing, the process was optimized so that more plastics and resin can be incinerated. The optimized incineration system still showed good adaptability when the proportion of plastic in the waste composition over 60%. The research on miniaturization and mobile technology of incineration were continuously carried out for the small reactors or small nuclear facilities, so as to further improve the economy. Compared with the incineration facilities, the mobile incineration technology can reduce the floor space by more than 90% and cut the construction cost by more than 75% under the same capacity and meeting the emission requirements.
Modeling and Analysis of Two-Phase Distribution in Large-Scale Separate Heat Pipe
Authors: Kuang Yiwu , Jin Xin , Zhuan Rui , Xie Fang , Chen Kaidi , Han Fei , Chen Xiting , Wang Wen
Published: November 2022

Large-scale separate heat pipe is widely used in various applications, such as waste heat recovery, passive cooling of Spent Fuel Pool (SFP) and so on. The evaporator and condenser of the heat pipe can be installed in different places with long distance. Working fluid circulates in the heat pipe due to static pressure difference between the evaporator and condenser. The separate heat pipe is a high efficiency heat transfer device which can transport heat for a long distance without considerable losses. A numerical model of separate heat pipe is established and verified according to the experimental data. The model agrees reasonably well with experimental data. The Mean Absolute Error (MAE) is 7.22%. According to the model, vapor quality at the evaporator exit in the separate heat pipe is analyzed. It is found that, vapor quality varies with heat pipe filling ratio and thermal loads. At low filling ratios, fluid at evaporator exit is overheated. While at high ones, it is in two-phase state. Vapor quality at evaporator exit decreases with the filling ratio. When the vapor quality is bit lower than 1, separate heat pipe comes to its maximum heat transfer capacity. It is also found that there is a minimum condenser height to activate the large-scale separate heat pipe. When the downcomer is full of liquid, elevating the condenser can increase the heat pipe performance. But when the condenser is high enough, the downcomer will be partially liquid filled, elevation of condenser provides little benefit.
Nuclear Security Against Insider Smuggling: Review and Recommendations
Authors: Al-Khodire Hiba
Published: November 2022

Smuggling of nuclear material is one of the serious problems of nuclear security. Nuclear institutes and organizations put great efforts to protect these materials from dangerous hands by applying nuclear security policies, using monitoring tools, and developing access control systems. These techniques effectively prevent outsider threats but cannot provide high confidence withstanding insider threats.

History offers many examples of violent employees, who broke the rules and smuggled nuclear material, taking advantage of their knowledge, experience, and authority to access the nuclear facility systems and their vulnerability. Nuclear smuggling operations can be through one insider, multi-insiders, or by the incorporation with an outsider, this makes the facility’s security planning a challenge.

The analysis of smuggling crime shows that the motivations for performing insider threats vary. It includes but is not limited to financial gain, poor security culture, and revenge. Lessons learned from other insider crimes in non-nuclear fields are useful for developing nuclear facility plans. Thus, it is enormously important to continuously improve the security system, overcome instrumentation malfunctions, and strengthen the nuclear security culture in the facility. This paper provides a review and recommendations in the legal and technical framework to genuinely prevent insider nuclear threats against nuclear facilities.
Parametric Study of Throttling Window Shapes on Flow Force Characteristics in Main Feed Water Regulating Valve
Authors: Zhao Lei , Ru Qiang , Wu Jia-Yi , Qian Jin-Yuan
Published: November 2022

As the key device of the secondary circuit system, the main feed water regulating valve plays a significant role in controlling water level, whose reliability directly affects the safety of the nuclear power plant. During the use of the main feed water regulating valve, the movement and stress state of the valve core are affected by the hydraulic force and the hydraulic torque, which is related to the accuracy and stability of the valve. In this paper, a typical throttling window type of main feed water regulating valve was taken as the research object, the shape coefficient of the throttling window was defined, and the effect of the throttling window shape on the force characteristics of the valve core under variable opening was analyzed. The results show that as the valve opening increases, the direction of valve core axial hydraulic force changes. And with the increase of the shape coefficient of the throttling window, the maximum axial hydraulic force and the maximum lateral hydraulic force of valve core first decrease and then increase, and the maximum hydraulic torque increases gradually. This study can provide some references for the optimal design of similar main feed water regulating valves.
Numerical Simulation of Single-Layer Wire-Mesh Sensor
Authors: Zhang Hengwei , Xiao Yao
Published: November 2022

Wire Mesh Sensor (WMS) is an advanced two-phase flow measurement device, which has been widely applied in experimental research and industrial production. Previous research has shown that the measurement uncertainty of the WMS increases with the increase of the axial distance between the two layers of the electrode wires. Moreover, the structure of dual layers of wire mesh limits the application of WMS in complex flow channels. For example, the structure of the rod bundle channel needs to be carefully designed to install WMS. Thus, a novel WMS with only a single layer of electrode mesh was proposed. The transmitter wires and the receiver wires were intertwined with each other to increase the stability. The signal response of the single-layer WMS was investigated by the potential field simulation. For the bubble with a diameter of two times the mesh size passing through the WMS, the signal response of the single-layer WMS is closer to the ideal signal compared with the two-layer WMS. For the bubble with a diameter of one time the mesh size, the single-layer WMS overestimates the void fraction. Near the intersection, the distance between the receiver and the transmitter is small, more currents would be received by the per length receiver wire. When the bubble contacts the single-layer WMS, it first wraps the highly sensitive wires, which leads to the void fraction being overestimated.
Research on Liability Insurance for Radioactive Sources and Suggestions to China
Authors: Li Xiao Ding , Gong Yu , Li Guang Hui , Bai Gang , Zheng Jie Ying , Qu Yun Huan , Li Zhe
Published: November 2022

Radioactive source liability insurance refers to the form of insurance in which the insurer shall be liable for compensation according to law when the insured uses radioactive sources to cause radiation damage or environmental pollution to a third party and is claimed for compensation. This paper deeply investigates the practical experience of various countries in the field of radioactive source liability insurance; this type of insurance has a lot of mature application experience in Europe & US, the two major nuclear damage compensation convention systems have not yet stipulated the liability insurance of radioactive sources, European countries have a strong awareness of applying for radioactive source liability insurance. Then it makes a detailed analysis of the current laws and regulations in the field of nuclear technology utilization and the current situation of liability insurance in the field of radioactive sources in China. At present, there is no special law on liability insurance in the field of radioactive sources, there are still several problems, including the weak awareness of risk prevention of enterprises using nuclear technology, the lack of social responsibility of some enterprises and their unwillingness to buy insurance.

In order to solve the above problems, the following four recommendations are made. First, improve the legal and policy system of radioactive source liability insurance in China. Second, pilot the implementation of radioactive source compulsory liability insurance in high-risk industries or enterprises. Third, promote personalized insurance rates and insurance limits. Fourth, carry out health damage insurance for employees in the field of radioactive sources at an appropriate time. Finally, it provides a strong and valuable reference for the future radioactive source liability insurance in China.
Functional Reliability Investigation on Capacity of Reactor Core Natural Circulation Cooling in XAPR Under MLOCA
Authors: Wang Baosheng , Ma Tengyue , Bao Lihong , Ma Zhenhui , Hu Pan , Li Da , Su Chunlei , Chen Lixin , Tang Xiuhuan
Published: November 2022

Due to the epistemic uncertainties involved in the numerical values of its parameters and its modeling, the system may not accomplish its function as required. Functional failure becomes an important factors in the natural circulation system operation failure, which needs to be considered in the system reliability analysis. In order to solve the problem of multi-dimensional uncertainty parameters and small functional failure probability of passive systems, an efficient functional reliability estimation method based on improved response surface and importance sampling subset simulation was presented. This method was applied in natural circulation cooling in XAPR reactor core. Combined with MLOCA, the uncertainties related to the input parameters and the model were considered. And then the probability of functional failure and the sensitivity measures were estimated with the combination of response surface method and importance sampling subset simulation method. The assessment results show that the functional failure probability is 3.796 × 10−3, and it is fully considered the functional failure in the passive system reliability analysis. Meanwhile, this method has high computed efficiency and excellent computed accuracy compared with traditional probability analysis methods, and it is robust to nonlinear functional function of XAPR.