bluuebunny/crossref_metadata_2025 huggingface dataset:
Effects of the Tip Clearance Size on the Tip Leakage Flow in an Axial Compressor Operating With Supercritical Carbon Dioxide
Authors: Gou Jinlan , Ma Can , Wu Jun , Li Bangming , Lin Yuansheng
Published: November 2022

The closed Brayton cycle operating with supercritical carbon dioxide is an attractive option for geothermal, nuclear and solar energy conversion. As one of the key component, the performance of the supercritical carbon dioxide compressor attracts lots of attention. Up to now, most studies focused on the centrifugal compressor and the design method of the axial compressor operating with supercritical carbon dioxide is far to matured. The effects of the tip clearance size on the tip leakage flow in an axial compressor are numerically investigated in the present work, with different tip clearance size and operating points. To fully take advantages of the matured design method of the air compressor, a similar air compressor is simulated as a comparison group. The results show that the flow characteristics especially the static pressure coefficient and tip leakage vortex structure is almost the same between S-CO2 compressor and the similar air compressor with small tip clearance size. This is also suitable for the large tip clearance size at low and middle span, while obvious discrepancy arises at high span. The tip leakage vortex of S-CO2 compressor is stronger and more downstream than the air compressor. Understanding these effects is helpful to extend the abundant experience of the air compressor to design S-CO2 compressor.
The Comprehensive Performance Comparison Research Between Two Types of Radiators Applied in Electronic Equipment
Authors: Huang Chonghai , Zhang Yelei , Yang Xiaohu , Ke Hanbing , Xiao Qi
Published: November 2022

The development direction of miniaturization and low noise of electronic equipment in nuclear power plant puts forward higher requirements for radiators. Two experimental systems were constructed to test the volume, noise and heat transfer performance of air-cooled radiator and micro-pump-driven water-cooled micro-channel radiator that can be applied to electronic equipment of nuclear power plant. The experimental results show that the comprehensive performance of micro-pump-driven water-cooled micro-channel radiator is better than that of the air-cooled radiator. Compared with the air-cooled radiator, the micro-pump-driven water-cooled micro-channel radiator can reduce the temperature rise of heat source by more than 20°C, and the volume is only 38% of the air-cooled radiator. For the micro-channel using in the present study, the flow rate of cooling water is the main factor which affects the heat transfer performance of radiator when the flow rate of cooling water is lower than 2.5L/min. However, when the flow rate of cooling water is more than 2.5L/min, the increase of cooling water flow rate has little impact on heat transfer performance, and the self-structure of the radiator and thermal resistance have become the main factors restricting the improvement of heat transfer performance.
The Effect of a Honey-Comb Porous Plate on the Critical Heat Flux Under Reduced Pressure Conditions
Authors: Wu Feifei , Mori Shoji
Published: November 2022

Enhancement of pool boiling heat transfer using a honeycomb porous plate (HPP) under atmosphere pressure has been experimentally examined. The previous research found that the critical heat flux (CHF) could be enhanced by up to three times above that of a plain surface. Considering the operating temperature in microelectronic devices, boiling at sub-atmospheric pressures for maintaining the lower surface temperature while removing high heat flux is required. Under circumstances of low pressure, there is a scarcity of research about boiling heat transfer for porous materials. Liquid supply is promoted by capillary action and vapor escape is facilitated by separating liquid flow channels from vapor flow channels in the HPP, and the CHF was expected to be improved even at reduced pressure conditions. Therefore, the CHF in saturated water boiling through an HPP at low pressure was studied in the current research. A more generalized understanding of the system pressure and the enhancement effect was provided, which made it possible to mitigate the enhancement technology bottlenecks through electronic devices and looked upon as further elaboration of the boiling heat transfer mechanism on honey-comb porous surfaces. There were three different pressures (51 kPa, 76 kPa and 100 kPa) investigated on the HPP; the results on the plain surface under the identical operating conditions were the same for each of the three pressures tested on the HPP.
Application of Probabilistic Safety Analysis in Analysis of Impact of Fangjiashan Nuclear Power Plant’s Load-Following Operation
Authors: Liu Deyi , Zha Sixing , Wu Jian , Cao Yong , Chen Honghao
Published: November 2022

At present, probabilistic safety analysis (PSA) has been widely used in many fields such as the design and manufacture of nuclear power plants, and PSA has gradually developed into an important tool for safety evaluation and decision-making. With the increasing peak-to-valley difference between power grid loads, the peak-shaving situation of the power system is becoming more and more severe. Due to the increase in the proportion of nuclear power in the power grid, the power system has an increasing demand for nuclear power units to participate in power grid peak-shaving. The nuclear power plant technical specifications and the nuclear power plant final safety analysis report in the nuclear safety regulations have clear requirements for the operating conditions of the nuclear power plant. As a power generation device with special safety requirements, the safest and most economical operation mode of nuclear power plants is to operate at rated load. The Chinese government and nuclear safety regulatory authorities attach great importance to nuclear power safety, and nuclear safety must be foolproof. From the perspective of probabilistic safety analysis, this paper analyzes the impact of Fangjiashan Nuclear Power Plant’s (FSJ NPP) participation in peak-shaving by using the load-following operation mode, and gives reasonable suggestions.
Investigation on the Flow and Heat Transfer Performance of Micro/Mini-Channel Liquid Cooling Heat Sink for Nuclear Control Chips
Authors: Yang Xiao-Hu , Zheng Zhaoli , Song Ping , Chen Lie , Huang Chonghai , Chen Kai
Published: November 2022

The “thermal barrier” problem of high-performance nuclear control chips is becoming more and more prominent with the increase of chip integration under the Moore’s Law, liquid cooling is an effective method to solve the problem of high power heat dissipation. This paper provides an analytical model of the flow and heat transfer performance of micro/mini-channel liquid cooling heat sink for high-end chips. The flow resistance model fully considers the friction loss along the way, the contraction and expansion pressure loss of the channel, and the local pressure loss at the entrance and exit; the thermal resistance model includes the spreading thermal resistance brought by the extension of the heat sink base, the conduction thermal resistance of the bottom plate, the convective thermal resistance between the coolant and the heat sink, and the capacity resistance of the coolant. Using this model, combined with numerical simulation, a typical liquid-cooling heat sink is analyzed, corresponding flow resistance and thermal resistance distribution characteristics are evaluated, and the optimal design configuration is given. The results in this paper can serve as valuable reference for the development of micro/mini-channel liquid cooling technology for high-performance chips.
Scaling and Engineering Design of Facility for Simulation of HTGR Reactor Building Response for a Break in Primary Coolant Break
Authors: Revankar Shripad T. , Kultgen Derek W.
Published: November 2022

Analysis and understanding the depressurization response of the HTGR reactor is essential to the understanding of accident scenarios and design of these reactors. The 350MWe GA-MHTR was used as reference reactor for the current design of the test facility. The design of the test facility was based on a well-established dynamic scaling analysis for integral system transient. The scaling accounts for integral system response, boundary mass and energy flow, system pressure, and local phenomena such as choking, convective and diffusive mixing, stratification, etc. The key phenomena determining the gas flow paths and distribution in the reactor building are depressurization, gas discharge, choked flow, venting, and mixing. Two scaled systems were developed, one 1/20 scale of the prototype and another 1/28 scale of the prototype. Based on sizes, the scale of 1/28 was chosen for the facility that would simulate all local and global phenomena in the blowdown process. The scaled 1/28 model design was then translated to engineering drawings to manufacture vessels and tanks for the test facility. The important aspect of the scaling/sizing the compartment lies in correctly scaling flow losses in the compartments, flow paths, and vent paths, gas hide-out spaces, and thermal stratification and mixing. Heat loss to the cavity walls is scaled via matching the thermal penetration depth/inertia of the model vessel (cavity) insulation material with that of the prototype concrete. The Engineering design considered the geometrical similarity to simulate the flow path and the hideout of oxygen in the reactor cavity.
Study for the Optimization of the Decommissioning Project of Nuclear Facilities
Authors: Iguchi Yukihiro , Kawasaki Daisuke , Yanagihara Satoshi
Published: November 2022

Basically, decommissioning of nuclear facilities is a project that does not generate new profit because it is carried out with the reserve funds from operation, etc. Therefore, its cost should be minimized with optimization by shortening the process and minimizing the waste, etc. Meeting the requirements of exposure risk (safety) also affects the optimization. In this study, we decided to integrate these evaluation methods to develop a comprehensive optimization evaluation method.

In this study, we established an average process for the current decommissioning plans of Japanese nuclear power plants and developed a cost evaluation method including sensitivity analysis.

As a result of examining the feasibility of the deferred dismantling strategy using the above calculation method, it became clear that although there is a reduction in disposal and dismantling costs due to the natural decay of radioactive materials, the maintenance and management costs during the safe storage period account for a large proportion of the costs, and for this reason, immediate dismantling is unconditionally advantageous, at least in Japan.

The components of optimization described above are naturally subject to various uncertainties and risks. For example, there are regulatory risks, and the location of waste disposal site is subject to social acceptance, so there is a great deal of uncertainty. In the future, these factors will be incorporated into the evaluation and studied, and the optimal strategy for decommissioning and what kind of uncertainty should be focused on will be clarified quantitatively.
Research on Modeling Method and Eigenvalue Calculation of Single Dispersion of Fuel Particles
Authors: Yuan Anmin , Wu Hongchun , Li Yunzhao , Cao Liangzhi , Wang Cong , Jiang Chaofei , Yu Shenzhi , Liao Yi , Qiu Jinrong
Published: November 2022

Dispersed particle fuel is an advanced form of fuel with many advantages, which is safer than traditional fuel. But a strong double heterogeneity effect is formed, due to the random dispersion of fuel particles in the matrix. The homogenization and uniform distribution approximation models will thus leads to severe errors. In order to account for the influence of the double heterogeneity effect, a homogenization model, a uniform dispersion model and a random dispersion model are constructed in this paper. Then the MCX program was used for calculation, and the fuels were compared and analyzed, subject to high fuel phase volume and different enrichment. The effect of particle diameter variation on the eigenvalues of the system is revealed, and the errors introduced by the model approximation are also evaluated. It is demonstrated by the results that the double heterogeneity is negligible when the diameter of the fuel particles is less than 100μm. However, the double inhomogeneous effect is relatively enhanced under large diameter fuel particles and low enrichment, which requires high-precision calculations. This paper has a certain reference significance for the evaluation of the double heterogeneity effect of the dispersed particle fuel and the correction of the traditional homogenization calculation.
Research on the Technology Accuracy of the On-Line Fracture Assessment of Cracked Pipe in NPPs
Authors: Chen Mingya , Yu Min , Gao Hongbo , Qi Shuang , Lin Lei , Xu Decheng , Wang Weiqiang , Shi Jinhua , Peng Qunjia , Zhang Guodong
Published: November 2022

The cracked pipes usually need to be replaced or conservatively evaluated in the Nuclear Power Plants (NPPs). There is still a lack of effective methods for a real-time crack propagation and safety evaluation based on massive monitoring transient data. From a technical standpoint, there is a need to establish inspection and monitoring technologies. This research has carried out a re-development in the general finite element analysis (FEA) software platform which is based on our research results and the requirements of the RSE-M Code. In this paper, the re-development is introduced firstly. Then, the information of the large pipe test information from France, Japan and China is summarized. Finally, the technology accuracy of the on-line fracture assessment of a cracked pipe in the NPPs is studied. The comparative research results show that the FEA analysis and modeling method of the pipe adopted by the authors in the re-development software is reliable and can meet the requirements for the on-line fracture assessment.
The Effect of Nozzle Safe End Hot/Cold Wire Tungsten Inert Gas Buttering on Microstructure and Property
Authors: Yang Chengdong , Zhang Maolong , Ru Xiangkun , Tang Weibao
Published: November 2022

Hot wire tungsten inert gas (HWTIG) and cold wire tungsten inert gas (CWTIG) have been used for nozzle safe-end buttering of nuclear island critical equipment, the effect of the HWTIG and CWTIG buttering on microstructure and property has been investigated. Research results revealed that the grain orientation of HWTIG buttering layer is mostly concentrated in (0 0 1) direction and the average grain size is 118.375μm, while the grain orientation of the CWTIG dispersed distribution in (0 0 1), (1 1 1) and (1 0 1) direction, and the grain size is 158.479μm. Furthermore, the grain boundary misorientation angle of HWTIG buttering layer has higher occurrence frequency between 10° and 30° high-angle boundary, whereas the CWTIG has higher occurrence frequency within 30° and 60°. Moreover, the microstructure of HWTIG buttering layer has less chromium-rich M23C6-type carbide compared to CWTIG, and the M23C6-type carbides discontinuous distribution on the grain boundary of HWTIG buttering layer. The buttering layer of HWTIG has lower ductility-dip cracking sensitivity than CWTIG.
Level 3 PSA Application in a UK Generic Design Assessment Project
Authors: Wang Jinkai , Wang Qi
Published: November 2022

Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice for L3 PSAs that have recently been performed worldwide and there is lack of mature standard to implement L3 PSA. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected as a primary tool for the accident consequence calculations. The strengths and limitations of the code are identified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results has supported the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and meet the UK regulatory expectation.
Effect of Cr6+ Concentrations on the Corrosion Behavior of 316LN Stainless Steel in Supercritical Water
Authors: Li Zhenhe , Huang Xi , Zhan Zixiong , Zhao Qi , Liu Junxiong , Wei Lihua , Li Xiaoyan
Published: November 2022

It is well known that the type and concentration of impurity ions in pressurized water reactor environments will significantly alter the corrosion behavior of materials. However, the effect of these water chemistry parameters on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. The type and concentration of impurity ions on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. In this study, the effect of Cr6+ concentrations on the corrosion behavior of 316LN stainless steel in 550 °C/25 MPa supercritical water (SCW) was investigated. The results show that the weight gain first decreased and subsequently increased with the increasing of Cr6+ concentration, which is 0.40±0.01 mg/cm2 for 0 ppm, 0.28±0.03 mg/cm2 for 0.1 ppm, 0.55±0.02 mg/cm2 for 1.0 ppm, and 0.59±0.01 mg/cm2 for 50 ppm, respectively. The Fe-rich magnetite oxides were gradually covered on the surface of samples, and its proportion on the surface followed a similar alteration trend as the weight gain, which is 44.3% for 0 ppm, 37.8% for 0.1 ppm, 67.7% for 1 ppm, and 87.6% for 50 ppm. A duplex structure of the oxide films, including an Fe-rich outer magnetite layer and a Cr-rich inner spinel layer, was developed on the regions covered with Fe-rich magnetite oxides. The alternation in the thickness of oxide film followed a similar trend as the weight gain, which is 7.67 μm for 0 ppm, 5.80 μm for 0.1 ppm, 9.97 μm for 1 ppm, and 11.23 μm for 50 ppm. However, a Cr-rich spinel oxide film was formed on the region without Ferich magnetite oxides. The thickness of this oxide film was unchanged regardless of Cr6+ ion concentration, which is about 2 μm, but the concentration of Cr in this oxide film is significantly higher than reference sample. These results indicate that the increasing of Cr6+ concentration in SCW will result in the corrosion rate to decrease first and then increase, which means that some measurements should be conducted during the operation of SCWR to keep a lower concentration of Cr6+.
Research on Source Characteristics of Radionuclide Diffusion in the Ocean Under Oceanic Radioactive Events
Authors: Li Zichao , Chen Rongchang , Liu Chen , Xue Qingqing , Wang Zhixia , Zhou Tao
Published: November 2022

All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of a nuclear leakage accident, radionuclides will diffuse through the atmosphere and the ocean, and pollute the ecological environment. To solve the problem of radionuclide diffusion in the ocean, we first need to solve the source term problem. The study on source characteristics of radionuclide diffusion in the ocean under oceanic radioactive events plays an important role in simulation and emergency decision-making. The meaning of source term and unit parameters were given, and types, forms and release modes of radionuclides were analyzed under nuclear leakage accidents. The storage and release of radionuclides were calculated. Then the dry and wet deposition fluxes of radionuclides were calculated. The results show that 131I, 134Cs and 137Cs are the most important radionuclides in radiation dose evaluation and they will be in the form of CSI and CsOH under nuclear leakage accidents. Radionuclides are released into the environment in the form of micron aerosol. Radionuclides entering the ocean include direct release and atmospheric radionuclide deposition. The dry deposition flux is 3 orders to 4 orders of magnitude lower than the ground radionuclide concentration. The wet deposition flux within 300m is much larger than the dry deposition flux, and the wet deposition flux outside 300m is about 1 order to 2 orders of magnitude larger than the dry deposition flux. It provides a prerequisite for the calculation of radionuclide diffusion in the ocean under oceanic radioactive events.
Research on Radionuclide Diffusion and Verification in the Ocean Under Nuclear Leakage Accidents
Authors: Li Zichao , Chen Rongchang , Xue Qingqing , Liu Chen , Wang Zhixia , Zhou Tao
Published: November 2022

All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of nuclear leakage accidents, radionuclides will diffuse through the ocean and pollute the ecological environment. The study of radionuclide diffusion and verification in the ocean plays an important role under nuclear leakage accidents. The hydrodynamic model and radionuclide diffusion model in coastal waters of nuclear power plant were established. The water level change and radionuclide diffusion area in coastal waters of nuclear power plant were calculated, and dangerous areas were given. The calculated results were verified by enteromorpha drift. The results show that radioactivity in the high concentration area can reach 109Bq/m3, and it is reduced by about 9 orders of magnitude to the source term. On the 14th day, the dangerous area is within about 58km to the east, 10km to the south and 21km to the west of the nuclear power plant. In the east-west direction, the dangerous area is about half of the radionuclide diffusion area.
Nuclear Power Plant Design Optimization Based on Risk-Informed Method
Authors: Li Wenjing , Liu Yu , Wang Gaopeng , Yu Xinli
Published: November 2022

The probabilistic safety analysis method is an effective tool to demonstrate the balance of design and identify the vulnerabilities. It can provide technical support for the design optimization of nuclear power plant. A nuclear power plant in China specially designed the reactor cavity injection system (RCI) with both active and passive trains to mitigate the severe accident with core damaged. When the core damage accident occurs, the melt debris in the lower head can be cooled by RCI with water injection to the outside surface of the reactor vessel, thereby maintaining the integrity of the vessel and realizing the core melt in-vessel retention (IVR). This paper will introduce the design of the cavity injection system, and use the PSA method to establish a model of the system. Then the reliability of the system and its contribution to the large release frequency will be discussed, the vulnerability of the system will be identified. System design optimization recommendations will be proposed based on PSA results, and the results comparison between the optimization scheme and the original design is discussed. It can be concluded that in the design process of NPP, risk-informed methods should be considered as an important technical support tool.
Flow Distribution Optimization Design of Certain Integrated Pool Lead-Bismuth Reactor
Authors: Wang Na , Zhou Youxin , Yuan Zhaojun , Liao Ziyu
Published: November 2022

The core inlet flow distribution of core is of great importance in the integrated pool lead-bismuth reactor design, while the integrated pool structure feature and lead-bismuth fluid bring difficulties to the flow distribution optimization. The result of flow distribution will directly affect the safety margin of thermal-hydraulic design: mal-distribution could lead to decrease of core thermal safety margin, which may cause negative effect on the reactor safe operation. Based on three-dimensional computational fluid dynamics (CFD) simulation, the flow distribution result is evaluated. It’s found that the eddy flow at the core entrance will lead to nonuniformity of flow distribution and flow fluctuation. Optimizations of design structure are conducted to improve the flow distribution characteristic. Different options of flow channel and flow distribution device design are proposed. By means of steady and transient calculation, the optimization options are evaluated from the two aspects: the uniformity and the stability of flow distribution. The local resistance loss and flow velocity are considered at the same time. At the end of the research, the structural design scheme which has better flow distribution characteristic and meets the restriction of local resistance loss and flow velocity is put forward. The research could provide reference to the thermal-hydraulic design of integrated pool lead-bismuth reactor.
Seismic Response Analysis of Pile Foundation of an Underground Storage Facility of Nuclear Power Project
Authors: Wang Dongyang , Sun Xiaoying , Liu Ziqiao , Gan Yingying
Published: November 2022

The proposed site condition of an underground storage facility of nuclear power project (NPP) is complex, characterized by strong spatial variability of bedrock surface, non-uniform soil layers and underlying soft soil. If the substructure method commonly used in nuclear power engineering seismic analysis is adopted, the non-homogeneous foundation must be simplified as the horizontal layered site condition. Therefore, the dynamic interaction features among piles, foundation soil and superstructures under strong earthquakes cannot be reflected accurately. Taking it as the project background, this paper conducts a research on the three-dimensional time-history seismic analysis method for NPPs. On the theoretical basis of viscoelastic artificial boundary and wave equation method, the soil-structure dynamic interaction analysis methodology in complicated non-homogeneous site is put forward in this paper, which is called the integral time-domain method. And an efficient platform is set up for computing artificial boundary parameters and seismic motions. Then the seismic response analysis of pile foundation of the underground storage facility is completed. It is concluded that the pile-soil-structure interaction does not affect horizontal response of the super structure significantly but leads to a frequency shift of the response peak in vertical direction. Additionally, the substructure method is used to verify the accuracy of the proposed method and the results of two methods match well. In conclusion, the integral time-domain method contributes to appropriately simulating the foundation radiation damping and seismic input. It has high accuracy and applicability. The research method and the computing platform can be taken as technical references for seismic response analysis of NPPs constructed in non-horizontal layered foundation or complex geologic bodies conditions.
Research on a Circuit Health Assessment Technology Combining Mechanism Knowledge and Data Driven Technology
Authors: Shi Guilian , Xiaojin Huang , Changyu Mo
Published: November 2022

Degradation of electronic equipment is typically accompanied by a deviation in their electrical parameters from their design values. When the function of the electronic equipment fails, parametric faults of the electrical parameters will ultimately occur. Such electrical parameters can characterize the state of health can be defined as health parameters. For complex electronic equipment such as hardware circuits of Nuclear power plant DCS cards, the use of health parameters for fault diagnostic has been widely used, but there is no mature method for sub-health diagnostic. To address this problem, an electronic equipment health assessment method was developed using a technology based on the combination of mechanism knowledge and data-driven. The method first uses Physics of Failure approach and hardware circuit logic principle to identify the major failure mechanisms of the circuit, and then determines the key health parameters and their sub-health thresholds on this basis. The time series data under specific test conditions for the key health parameters of the circuits to be assessed is collected and compared with the sub-health thresholds, then the health status assessment results based on mechanism knowledge can be obtained. As well, the data-driven health assessment technology based on Mahalanobis distance is applied to analyze the real-time health parameter data of all samples, and the healthy outlier samples are identified and quantified, and the health status assessment results based on the data-driven technology can be obtained. The results of the two are comprehensively evaluated to finally determine the circuit health status. Practice was carried out on the DCDC chip of a nuclear power DCS analog output module, and the results showed that this method was feasible for engineering.
HL1000 Nuclear Power Station Emergency Borating System Capacity Demonstration
Authors: Tian Zhaohua , Li Qiang
Published: November 2022

Emergency Borating System has been widely used in the new third generation nuclear power plant in the world. HL1000 nuclear power plant is equipped with the Emergency Borating System (RBS) as an engineered safety facility, which provides the reactor the high pressure safety boron injection and long-term boration functions to ensure the core in subcritical. Firstly, the main functional requirements of RBS system of new third generation nuclear power plants in the world are introduced. Then, according to the two functional requirements for RBS of HL1000 nuclear power plant, reactivity control and confinements, the design criteria for RBS are determined. Based on these design criteria, the methodologies for RBS capacity demonstration are developed by the multi status point method. The relevant capacity parameters of RBS, including the maximum and minimum injection flow of RBS and the minimum volume requirements of boric acid tank are calculated separately.
Study on Interfacing System LOCA of VVER-1200
Authors: Du Jinyan , Dong Fangyu , Ma Chao
Published: November 2022

The evaluation of public risk at nuclear power plants has generally found the risk to be very low. Part of the reason for the low assessed risk can be attributed to the “defense-in-depth” philosophy of nuclear power plants. This defense-in-depth includes multiple barriers to prevent the release of radionuclides to the environment. Nevertheless, accident scenarios have been postulated which might compromise the defense-in-depth by bypassing the reactor coolant system boundary and the containment. One such accident type is referred to as the interfacing system LOCA (ISLOCA). This type of sequence involves the loss of isolation between the high pressure RCS and a low pressure system outside containment. ISLOCA could lead to the loss of coolant outside containment while simultaneously disabling ECCS injection. Such a scenario could result in core damage and might have substantially higher consequences than some of other postulated severe accidents because the containment is also bypassed. Thus, it is necessary to calculate the ISLOCA risk for a specific nuclear power plant. Based on summarizing the ISLOCA analysis method and process, this article carried out the ISLOCA analysis for the VVER-1200 nuclear power plant, identified its typical ISLOCA path, and quantified the impact on the safety of nuclear power plants.
Automatic Scram Forecasting During Abnormal Conditions of Nuclear Power Plants Based on LSTM
Authors: Chen Hanying , He Xiaoqiang , Gao Puzhen , Tan Sichao , Hou Xiaofan
Published: November 2022

During abnormal conditions of nuclear power plants that do not lead to automatic scram immediately, operators need to monitor the transient condition of nuclear power plant in order to evaluate the development rate of the abnormal conditions. Conventional condition monitoring methods cannot provide operators with an alert or forecast for automatic scram. In this work, a deep-learning model is presented for predicting the automatic scram assuming no operator intervention. The presented prediction model is based on Long Short-Term Memory (LSTM) model, which is a special type of gated Recurrent Neural Network (RNN) designed to adaptive control the memory length of the learned features and showing good performance in Sequence to Sequence (Seq2Seq) problems. The predicting model is trained by transient data of abnormal conditions: the input of model is the nuclear power plant monitoring data; and the output is the remaining time from current time to reactor trip. The predicted remaining time to reactor trip decreases with the development of abnormal condition, thus the output of prediction model will generate a countdown to reactor trip. Prediction experiments were conducted on simulated abnormal condition data generated by a full-scale simulator for a 300MW pressurized water reactor (PWR). The data with artificial noise were also used to test the performance of the presented model. The experiment results shows that the presented prediction model has better prediction performance and is more robust to noisy input than models based on one-hidden-layer feedforward neural networks or simple RNNs (e.g., Elman network). The presented intelligent prediction model can provide nuclear power plant operators with a countdown alert about the upcoming reactor scram, as such the operators can prepare or take countermeasures for it.
Simulation on the Multidimensional Behavior of TRISO Particle Fuel With UN/ UO2 Kernel
Authors: Wu Changwei , Wu Junmei , He Yanan , Wu Yingwei
Published: November 2022

As one of the new advanced fuels proposed to improve the safety of nuclear reactors, fully ceramic microencapsulated (FCM) fuel consisting of TRistructural ISOtropic (TRISO) particles dispersed in a silicon carbide matrix has received a lot of attention at home and abroad because of its distinguished advantages such as high corrosion resistance and excellent fuel irradiation stability. For analyzing FCM fuel performance and optimizing FCM fuel design, it is necessary to assess complicated behaviors of TRISO particle fuel under irradiation. To investigate the multidimensional behavior of TRISO particle fuel with UN or UO2 kernel, a three-dimensional thermomechanical coupling model was developed based on multiphysics coupling finite element software COMSOL. Additionally, thermomechanical performance analysis of TRISO particle fuel with irradiation-dependent material properties and behavior models was conducted, and important phenomena were implemented such as fission gas release, production and diffusion of CO gas, as well as the evolution of particle radius due to the irradiation deformation, fission product swelling strain and irradiation creep. Moreover, on the basis of the preliminary thermomechanical analysis, the performance of TRISO particle fuel with UO2 and UN kernel were compared. This paper is intended as an exploratory investigation of TRISO fuel behavior, which will provide vital reference for further performance analysis and design optimization of subsequent FCM fuel.
Design and Commissioning of Digital Control System for Refueling Machine in HPR1000
Authors: He Zhijun , Zhang Peng
Published: November 2022

The design of digital control system for refueling machine in HPR1000 fully absorbs and refers to the experience of independent design and supply of refueling machine in the previous PWR projects, the experience of technical service in the operating nuclear power plants and the experience feedback from owners.

This paper describes the design scheme, improvement and commissioning of refueling machine digital control system for HPR1000 in detail. It focuses on the design of system architecture, hardware selection, PLC system configuration, PLC program development and HMI configuration, etc. At the same time, in order to reduce the work intensity of operators, the digital control system enhances the function of HMI, improves the display of fault information, and realizes the automatic digital control of refueling machine. In HPR1000 reactor, the digital control system of refueling machine has been optimized and improved in both hardware design and software design, which greatly improves the safety, reliability and economy of refueling machine.
Experimental Study on Measurement of Boiling Entrainment Rate From Falling Liquid Film With Gas Sheared Flow
Authors: Permana Raka Firman Baskara , Tabuchi Junpei , Okawa Tomio , Narushima Yuki , Furuichi Hajime , Katono Kenichi
Published: November 2022

Study on the droplet entrainment in annular two-phase flow has attracted many researchers in connection with the liquid film dryout in the annular two-phase flow. However, the knowledge on the entrainment due to nucleate boiling is still insufficient to predict the droplet entrainment rate accurately. In this study, measurements were made of the droplet entrainment rate with falling liquid films on the locally heated test section. The visualization using a high-speed camera to clarify the accuracy of the measurement results also carried out in this experiment. For the first step in this study, the measurement of the droplet entrainment rate was performed after getting a stable or steady state conditions. Then, we investigated the influence of the separator position on the reproducibility of the droplet entrainment rate to get the droplet entrainment rate – separator position map. As a result, the reproducibility of the droplet entrainment rate increases with decreasing separation distance from the liquid film. In additions, we also investigated the influence of the heat flux and liquid film flow rate on the reproducibility of the droplet entrainment rate. As a result, the reproducibility of the droplet entrainment rate increases with increasing the heat flux and liquid film flow rate. In the future, we expect to investigate the influence of gas flow rate, and degree of subcooling on the reproducibility of the droplet entrainment rate and some correlation models are developed.
Optimization of MOX Fuel Assembly Based on Genetic Algorithm for Discussion on Plutonium Consumption
Authors: Yamamoto Yuma , Kitada Takanori , Takeda Satoshi , Takeishi Taichi
Published: November 2022

Japan has been promoting its Pu utilization program based on the principle of not having excess Pu from the viewpoint of non-proliferation of nuclear weapons. On the other hand, by the end of 2020, Japan had 46.1 tons of separated plutonium, and the Rokkasho reprocessing plant, which will extract separated plutonium from spent nuclear fuel, is scheduled to start operation. Currently, there is a lack of ways for Pu consumption, and the practical way of Pu consumption is the use of MOX fuel in light water reactors. However, the conventional nuclear design of MOX fuel assemblies of light water reactors is mainly determined by focusing on the local peaking factor, and the Pu consumption has not been considered much. In this study, we focused on the Pu consumption in the MOX fuel assembly and obtained the fuel pin arrangement that maximizes the Pu consumption by changing the fuel pin arrangement and Pu composition using a genetic algorithm. Focusing only on the local peaking factor, the optimization was performed to obtain the fuel pin arrangement that minimizes the local peaking factor. As a result, it was found that fuel pins with high enrichment were placed in the fuel assembly where Pu consumption was maximized, except for the fuel pins adjacent to the uranium fuel where the thermal neutron influx has a large effect. In the arrangement that minimized the local peaking factor, the local peaking factor is suppressed to 1.1. From these results, it was confirmed that the optimization can increase the amount of Pu consumption.
Molten Salt Reactor Experiment Simulation Using ThorMOC
Authors: Zhang Ao , Dai Ming , Cheng Maosong , Chen Jingen
Published: November 2022

High fidelity modeling and simulation of nuclear reactors with direct whole core transport calculation has become an important state of art for development of reactor computational tools with the rapidly improvement of computing power. In this work, a quasi-2D delayed neutron precursor (DNP) drift model has been implemented in a GPU-based whole core transport code ThorMOC to achieve high fidelity modeling and simulation of liquid-fueled molten salt reactors (MSRs). A rasterized coarse mesh finite difference (RCMFD) method is applied in ThorMOC to improve the convergence rate, and to handle complex MSR models with arbitrary boundaries by combining with long characteristics method. OpenMC is used to generate a multi-group cross section library and to provide reference results for stationary state calculations. The results including eigenvalue and power distributions from ThorMOC agree well with those from OpenMC for stationary state calculations of the Molten Salt Reactor Experiment (MSRE) model. The reactivity loss due to DNP drift is computed and is in a good agreement with the measured value in experiment.
Experience Feedback Report on Hydrogen Meter Discontinuity of Steam Generator Overflow Tube
Authors: Liu-Huan Zhang , Xiang Wu , Dong-Fang Ouyang
Published: November 2022

The hydrogen meter sampling branch of 1st ring steam generator overflow pipe of CEFR secondary loop from 2016 to March 2020 has no access to sodium flow. The unavailability affects the way judges the small leakage in the upper part of the steam generator. After investigation, in March 2020, the sampling branch restored sodium flow, and the hydrogen meter returned to normal. In this paper, the cause analysis, troubleshooting process and solutions of hydrogen meter sampling branch are introduced, and the operation and maintenance experience of diffused hydrogen meter system is summarized, providing reference for the design, optimization, maintenance and personnel training of diffusion hydrogen meter in China experimental fast reactor (CEFR) and demonstration fast reactor.
Mechanical Performance Assessment of Independently-Developed Zirconium Alloy Under RIA Condition
Authors: Wang Jie , Wang Yang , Wei Jun , Deng Yongjun
Published: November 2022

Reactivity-initiated accident (RIA) is postulated design-basis accidents (DBAs) in light-water reactor (LWR). Moreover, Pellet-cladding mechanical interaction (PCMI) can cause a failure of the cladding in the early transient. In recent RIA regulatory guide released by NRC which points out that PCMI failure threshold depends on total hydrogen content in cladding during RIA. In order to evaluate the performance of independently-developed zirconium alloy under RIA conditions, the mechanical behavior and the fracture of 2 different cladding tubes (CZ, and SR (Stress Relief) Zr-4) with different hydrogen contents are investigated under thermal-mechanical loading conditions representative of PCMI during RIAs. Ring tensile tests are performed at room temperature, 350 °C on 2 different materials containing various hydrogen concentrations up to 1000 wt. ppm. Test results indicate that the ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking, the ductility and strength results of SR Zr-4 have a good agreement with reference paper, confirming the rationality of experimental method applied and reliability of test facilities. According to the results of Zr-4 and CZ, a conclusion can be made is that the ductility of independently-developed zirconium alloy (CZ) is better than Zr-4, which can provide the technical support when licensing.
Comprehensive Comparison and Analysis for Fast Reactors
Authors: Yan Zheng , Zhao Weiguang , Cao Xiaxin , Yan Changqi
Published: November 2022

Fast reactor is the main reactor type of the fourth generation reactors. It has the advantages of good fuel breeding characteristics, strong transmutation ability, high thermal efficiency and good safety. The research of fast reactor is the focus of current advanced reactor. According to the characteristics of fast reactors, the differences between fast reactors and pressurized water reactors in the fuel breeding, transmutation and safety are analyzed in this paper. On this basis, the characteristics of sodium-cooled fast reactors, lead and lead-bismuth eutectic cooled fast reactors and gas-cooled fast reactors are analyzed and their feature differences are compared. According to the basic characteristics of reactor physics, thermal-hydraulics and safety, the characteristics of these fast reactors are analyzed and compared, their respective advantages are given, the technical problems and difficulties are pointed out. The advantages and suitable application fields of these fast reactors are illustrated, the development opportunities and technical challenges are discussed. Some analysis results can provide reference for the research, design and application of advanced nuclear reactors.
Design Verification of Minimum Flow Valve of the Main Feed Water Pump Based on the Full-Range Simulation Platform of CPR 1000
Authors: Zhang Zhenhua , Zhang Bo , Zheng Chaoying
Published: November 2022

When the flow of the main feed water pump of CPR1000 unit is low, that is, when the unit is operating at low load, in order to protect the main feed water pump from damage, two low-flow protection recirculation pipelines are installed at the pressure stage pump outlet, and the two lines are installed with two minimum flow regulating valves. These two valves generally have two designs: proportional type and on-off type. The paper simulates the response of the minimum flow valve of main feed pump to water level regulation of steam generator at low load based on the full-range simulation platform of CPR1000 nuclear power unit. By analyzing the response curve of the water level of the steam generator, the advantages and disadvantages of the different valve designs are verified separately. In the proportional design, the valve position of the minimum flow valve is easy to oscillate with the fluctuation of the flow signal, which will affect the stability of the steam generator feed water control system and then affect the operation of the water level regulation system of the steam generator. At the same time, the proportional minimum flow control valve has the control dead zone. Under the transient conditions with frequent disturbance, the response speed of the minimum flow valve does not match the response of the water level adjustment of the steam generator, resulting in the unstable water level of the steam generator. In the on-off design, the stability of water level regulation will have certain advantages. After comparing the influence of the two valve designs on the water level regulation of the steam generator and analyzing the limitation of the proportional design and the importance of minimum flow valve, two improvements schemes are given for the design of the minimum continuous regulating flow valve.