bluuebunny/crossref_metadata_2025 huggingface dataset:
Research on In-Core Reactivity Control Method and Neutron Characteristic Analysis of a Space Gas-Cooled Reactor
Authors: Chen Zhiqi , Cai Jiejin , Li Xuezhong
Published: November 2022

Compared with the conventional power system, the space nuclear reactor power system can provide a more stable and longer-life electricity supply, which can better meet the requirements of deep space exploration. For space gas-cooled reactors, the impact of the submerged accidents after a launch abort situation must be considered. In such an accident, the core void is filled with seawater and wet sand, the neutron spectrum in the reactor is thermalized, and the reactivity increases significantly and probably making the reactor supercritical. In this paper, the research focuses on a novel control system named the control ring. Two different neutron-absorption materials are doped in the control ring, and the reactivity control effect of the control ring in submerged accidents is analyzed with different enrichments of fuel. Also, the burnup and neutron spectrum of these two materials are also studied. The result shows that the performance of the reactivity control effect of gadolinium in the submerged accidents is equal to boron-10, and the control effect of the control drums will slightly increase when gadolinium is doped in the control ring. In addition, the reactivity penalty of gadolinium is smaller than boron-10. These results can provide a reference for the design of space gas-cooled nuclear reactors.
Effect of Contact Stress on Fretting Corrosion Behavior of Alloy 690 Tube Exposed to Simulated Secondary Water
Authors: Zhuang Wenhua , Han Zhongli , Guo Xianglong , Zhang Lefu
Published: November 2022

Flow-induced vibration is unavoidable in the steam generator, so the wear of heat transfer tubes can’t be ignored. It is essential to evaluate the fretting wear damage of the heat exchanger tube. In this paper, the high frequency fretting wear tests of Alloy 690 tubes/405 stainless steel plate were carried out in the dynamic water loop experimental device, simulated the secondary water (285°C, 10.5 MPa). Changing the contact stress is realized by applying a range of normal forces. As the contact stress increased, the wear volume increased, and the material removal increased the contact area and wear depth. The wear coefficients of Alloy 690 tubes were found to be intensely dependent on the contact stress. The damage mechanisms of Alloy 690 tube were mainly delamination and abrasive wear, and accompanied by material oxidation in the fretting process. The oxide type of wear debris mainly consisted of (Ni, Fe) (Fe, Cr)2O4 and Fe3O4. The profile shape of wear scar was dependent on the action of the wear particles. When the contact stress is minor, the debris was easy to escape from the interface, and the typical profile shapes were U-shaped. As the contact stress increased, the elastic deformation of the contact material increased, followed by adhesion and delamination in the contact area.
Application of Stiffness Confinement Method on Pin Resolved Variational Nodal Method and Its Implementation to C5G7-TD Benchmark Problem
Authors: Xiao Wei , Zhang Tengfei , Liu Xiaojing , He Donghao , Pan Qingquan
Published: November 2022

This paper presents an implementation of the stiffness confinement method in the framework of the variational nodal method (VNM) to solve time-dependent pin resolved neutron transport equations. With frequency-transform, the time-dependent neutron transport equation is decoupled from the precursor equation and transformed into a transient eigenvalue problem (TEVP) which can be solved with the in-house VNM code VITAS-FE. The verification tests were performed using the C5G7-TD benchmark problems. The result agrees well with solutions of the MPACT. The study indicates that the VNM and SCM can be applied to solve the high-fidelity pin-resolved transient problem accurately.
The Study on Microwave Process for Spent Resin
Authors: Chao Gao , Meilan Jia , Hongxiang An
Published: November 2022

The moisture content of spent resin from NPPs is generally 50% ∼ 60%, which concentrates radionuclides such as 60Co, 90Sr and 137Cs. It belongs to wet solid waste, accounting for about 1/4 of the total solid waste of NPPs. Pent resin is a dispersive material, which must be stabilized and packaged to meet the requirements of waste transportation and final disposal; However, the characteristics of radioactive organic waste also cause the radiolysis of spent resin and the production of H2, CH4, NH3 and other gases; The spent resin contains more sulfur and nitrogen, and the incineration products and degradation products have strong corrosive to the equipment and storage containers. Therefore, how to treat the spent resin safely and effectively is a difficult problem.Microwave process for spent resin using microwave penetration ability and body heat characteristics, remove the water and organic components from spent resin, so as to achieve the inorganic for organic waste. Compared with the original resin, the ashing product capacity and weight reduction ratio greatly. The spent resin dried by microwave does not contain free water, the volume is 1/3 of the resin volume before drying, and the mass is half of the resin mass before drying; After Microwave ashing, the ashing product of the spent resin does not contain organic components, and the ashing product is only 1/20 of the mass of the dried resin and 1/3 of the volume of the dried resin, that is, for the spent resin with a moisture content of about 50%, after microwave drying and ashing process, not only the organic components in the spent resin are removed, but also the volume and quality are greatly reduced. Most of the treated radionuclides exist in the ashing products, 60Co in the ashing products is 1011 Bq/kg and 137Cs is 108 Bq/kg. The technology processing are used to get the product, which can be convenient for subsequent processing. On the study, the technological parameters of microwave drum drying and microwave ashing of spent resin are determined.
Analysis of Trace Components in Neon by Gas Chromatography
Authors: Yang Li-ling , Zhan Qin
Published: November 2022

In order to complete the test of trace concentrations in neon carrier gas of solid tritium breeder system for fusion reactor, it is very necessary to establish a high precision analytical system and develop a method of Gas Chromatography (GC). The GC system was composed of three detectors and five separated columns and other auxiliary systems, meanwhile it has established analysis methods of testing trace He, H2 and impurity components in Ne carrier gas. The results showed that the Relative Standard Deviation (RSD) values of the concentration and peak area of each component were less than 5.0%, and the linear correlation coefficients (R2) were greater than 0.99 with the GC analysis system, which could be used to testing trace concentrations in neon with good repeatability. It could support the requirement of tritium efficiency in the Tritium Analysis and Measurement system (TAMS), besides, it could provide technology data and support for the radiation tritium system.
A Modified Two-Zone Model for Predicting Heat Release Rate of Pool Fire in a Confined Space
Authors: Peng Lan , Huang Xianjia , Wang Jinkai , Zhu He , Yang Ping , Xing Chaoliang , Zhao Chunyang
Published: November 2022

The heat release rate of the fire source is one of the most important parameters for the fire hazards analysis in nuclear power plants. For a fire in a confined compartment, the oxygen concentration has a critical effect on the fire heat release rate under oxygen-deficient situation. In the present work, a modified two-zone fire model was developed to predict the pool fire heat release rate in the oxygen-deficient situation. A simple model estimating the effect of the oxygen concentration on the fire heat release rate was incorporated into two-zone model, CFAST. Furthermore, the conservation of fuel mass was also taken into consideration. Experimental results from three fire experiments of pool fire under the oxygen-deficient situation in the literature available were used to validate the reliability of the modified two-zone model. The oxygen concentration in the compartment was decreased and became oxygen-deficient condition due to the mechanical ventilation. Compared with the original CFAST model, the accuracy of predicting pool fire under oxygen-deficient situation was increased considerably by the modified model, especially for the pool fire at steady under oxygen-deficient situation. In comparison with the experimental data, the cosine similarities of the three heat release rate curves were all over 0.95 and the max relative error of the burning time was 41.9%, which is significantly lower than that of the original two-zone model prediction.
Discussion on Setting Value and Grounding Mode of Grounding Protection for the Construction of a Nuclear Power Plant Power Loop Network
Authors: Yuhan Wang , Zhijian Wang
Published: November 2022

Based on the case of a single-phase ground fault protection failure in the power supply ring network of a certain project, through the calculation of capacitance current and the protection setting calculation, the reason for the failure of the protection is analyzed in depth, and further suggestions for protection settings are put forward. Finally, based on the above analysis and combined with the actual situation of the project, an in-depth discussion on the grounding method of the construction power ring network was carried out, and the direction of optimization and improvement was pointed out.
Discussion on Selection of Codes and Standards on Nuclear Power Plant Safety
Authors: Yuxiu Chen , Yuhong Wang , Haiying Luo , Jinkai Wang , Qing Shi , Zengpei Zhu , Bei Wu
Published: November 2022

Requirements of evaluating and determining applicability, adequacy and sufficiency of codes and standards applied in nuclear power plant design are required by IAEA, NNSA etc. without detailed guidance or suggestion yet. Purpose of the article is going to discuss and provide a methodology for selecting codes and standards methodology to evaluate their applicability, adequacy and sufficiency, purpose is tying to react those requirements. The methodology is explored based on GDA (Generic Design Assessment) practice, 5 steps are included: breakdown of design activities, collection and identification, evaluating applicability and adequacy, evaluating sufficiency, evaluating consistency. Process and detailed evaluation method are discussed. Selection process could supplement the gap of selecting applicable codes and standards, providing a mentality for engineers to focus on and continually raising other better process or methods for ensuring applicable codes and standards can be selected at the beginning of the nuclear power plant design and implementing nuclear security culture more deeply on selection of codes and standards.
Simulation Analysis of Main and Auxiliary Switching and Motor Starting in Nuclear Power Plants
Authors: Xingzhen Zhang , Hui Sun
Published: November 2022

The main and auxiliary switching process involves the large-capacity asynchronous motor used in various systems (such as water pumps, fans, etc.) that are accompanied by strong electrical and mechanical shocks during the restarting or reinput process, especially the influence of the motor group restarting shock. The process includes the dynamic behavior of the electromagnetic and mechanical transient change process. In Order to quantitatively analyze the dynamic behavior of induction motor starting, the general steady-state analysis method can no longer be accurately solved, it is necessary to consider the dynamic load model of the transient process, and select ETAP software to realize the model establishment and transient process simulation calculation. For high-power motor load starting, the motor may fail to start due to low starting torque or other running motors connected to the power grid. Therefore, this paper selects typical high-power motors as the research object as the largest single load, and considers the harsh power supply conditions, and the transient conditions of the above-mentioned plant power switch motor group start, To verify whether the motor starting or restart will cause itself and other motors to stall, or whether it causes the EDG diesel generator to start. Whether the power system of nuclear power plant can resume stable operation after the above transient process has become an important index for evaluating the power system design of nuclear power plant. In this paper, the simulation study of the switching between the main power supply and the auxiliary power supply of the nuclear power plant, and the starting process of the high-power motor is established, the power system model of the power grid is analyzed at high voltage and low voltage, and the transformer is automatically loaded. The on-load voltage regulation, the plant power switching process and the high-power motor starting simulation calculation, effectively reduce the cost of the experiment, Has great economic benefits, it has positive significance for fault analysis.
Study on Influencing Factors of Soil Sample Self-Absorption With Monte Carlo Method
Authors: Ye Erlei , Shen Chunxia , Nan Hongjie
Published: November 2022

The detection efficiency of soil samples in a cylindrical measuring geometry was calculated using the Monte Carlo method, evaluating the self-absorption corrections in the energy range of 46-2615 keV. By controlling variables, the effects of parameters such as sample density, height and humidity on the self-absorption factor have been studied, and the corresponding correction functions have been obtained. The research results show that: for γ photons of a specific energy, the change of sample density has the greatest impact on self-absorption. For samples whose density is not much different from that of the standard sample, the impact of changes in height and humidity on self-absorption should be considered. In the high-precision measurement of samples containing low-energy γ-photon radionuclides, the errors caused by density, altitude and humidity should be comprehensively considered.
Research on Intelligent Classification Method of Primary Loop Transient in Nuclear Power Plant
Authors: Zhang Run Ze , Chen Zhi Lin , Huang Yong Qiang , Chen Xin
Published: November 2022

The primary circuit transient of nuclear power plant is one of the important references for characterizing the fatigue of primary circuit materials. For the classification of transients, the commonly used method is to determine the category of transients according to the change in value and the operating operation of the unit. This paper forms a complete set of digital transient characteristics through the study of the thermal parameters behind various transients and the exploration of the law of transient occurrence. Combined with new computer technologies such as machine learning, the automatic processing of transient classification is realized. It provides ideas for the key links in the development of the whole-process automation platform for transient management.
Preliminary Sensitivity and Uncertainty Analysis of Accident Tolerant Fuel in SMR
Authors: Liang Yuxuan , Wang Xiang
Published: November 2022

As the technology of small modular reactor (SMR) becomes mature, its safety is getting more attention. In recent research of accident tolerant fuel (ATF) such as the U3Si2 used in a large commercial pressurized water reactor can significantly improve the safety of the core due to its high thermal conductivity and good physicochemical properties. Therefore, applying the ATF in the SMR can also enhance its safety. The mPower with a thermal power output of 530 MW designed by B&W company is selected as one of the SMR. In the design and production process of fuel rods, the design and manufacturing parameters such as fuel pellet diameter, inner diameter of cladding, fuel pellet density and other parameters have uncertain factors. These factors affect the results of effective multiplication factor (keff). In this work, we carried out preliminary sensitivity and uncertainty analysis of the U3Si2 with accident tolerant cladding (FeCrAl, SiC) in mPower using SEPRENT and DAKOTA. Sensitivity analysis performed the relative importance of each input parameters for different ATF combination. Uncertainty analysis gave the tolerance range and uncertainty for different ATF combination.
Construction and Application of Collaborative Editing System for the Nuclear Power Large Comprehensive Report
Authors: Luo Cheng , Jia Lin , Hao Wang , Wu Xiangyong
Published: November 2022

The nuclear power large comprehensive reports need a lot of editing work and participants, the process is opaque, the format is not standardized and there are no effective collaborative tools. Because of overall process improving requirements, report review needs of format verification, optimization process, centralized storage of documents, and specification of document content, our company is in urgent need of new technical means. Collaborative editing is a new technology that is fundamental and critical in multidimensional collaboration of organizations, tasks, resources, processes, and format specifications. This thesis is based on the promotion of collaborative editing of large comprehensive reports, in-depth study of related technologies including the timeliness of collaborative editing, unified document format standards, unified and effective process control, unified document storage and version management from tools, processes and platforms, and proposes a construction plan to solve these existing problems, and finally provides a collaborative editing platform for each major. Construction and application of collaborative editing system for the nuclear power large comprehensive report realizes workflow and standardization, greatly improves editing efficiency and quality of large comprehensive reports, reduces manpower input, human errors, saves management costs of various professions and lately helps HPR1000 successfully pass qualification verification.
A Numerical Study on the Behaviors of Graphite Aerosol in the Containment During a Water Ingress Accident of High Temperature Gas-Cooled Reactors
Authors: Fang Zhu , Zhang Yiyang , Wang Yating , You Zhikai , Sun Libin , Wu Xinxin
Published: November 2022

The graphite dust is a special problem for pebble-bed high temperature gas-cooled reactors (HTGR), due to the close coupling with the fission products. A large fraction of graphite particles deposit on the tube bundles of steam generator (SG) due to the particle-vortex interaction and thermophoresis effect. Once a water-ingress accident occurs, the graphite dust may resuspend due to the rapid change of flow field. A certain fraction of resuspended dust, together with some fission products, could be discharged to the containment if the overpressure protection of the primary circuit is triggered. The distributions of deposited and airborne particles in the containment are crucial for the safety analysis and source term evaluation. In this work, we use a Eulerian-Lagrangian method to numerically study the transport and deposition behaviors of graphite aerosol in the full-scale containment during a water-ingress accident. Particularly, the effects of highly irregular shape of graphite particles on the particle-wall and particle-fluid interactions are incorporated. The statistical results that as the tracking time increases, the deposition rate first drops rapidly and then exponentially decreases. The deposition fraction is 80.38% for the floors, 9.83% for the shell of the containment, and 3.55% for other surfaces. Only 6.24% of particles are suspended in the air. As the particle size increases, the deposition fraction increases first and then decreases, reaching a maximum value at 3 μm. We also discuss the effect of the down direction of overpressure relief outlet on particle deposition. The result shows a fraction of 11.68% of airborne particles, which is about 2 times than that of the up direction. Our simulation not only demonstrates that the containment has a good capability to retain the graphite dust in a water-ingress accident scenario, but also provides a support for the engineering design of the overpressure relief system of HTGRs.
Radiographic Examination of Electric Heater Tube Weld for HPR1000 Pressurizer
Authors: Yu Jintao , Lu Dandan , Zhang Jianlei
Published: November 2022

HPR1000 (Hua-long Pressurized Reactor) nuclear power technology is a third generation nuclear power technology independently developed by China and has completely independent intellectual property rights. The pressurizer is one of the most important equipments in nuclear power plants. Compared to the regulator of the CPR1000 pressurizer, the HL-1000 pressurizer designed by CNPDC has as many as 108 electric heaters. Each electric heater contains a weld of electric heating sleeve to connection. RT should be completed at least 216 exposures, if each weld is exposure with a minimum of two 90 degree shots, in this paper, the authors have improved RT examination technique of these welds and verified it on the actual products. The results show that the new technique can reduce the number of exposures by more than one third and speed up the manufacture of the whole pressurizer regulator.
Numerical Simulation on the Flow-Induced Vibration of Two Square Cylinders in Staggered Arrangement
Authors: Wu Yidong , Shi Li , Wu Xinxin , Wang Xiaoxin
Published: November 2022

Based on the design of the heat transfer tube bundles with square cross section of the steam generators in the high temperature reactor-pebble bed modules (HTR-PM), the flow-induced vibration (FIV) of two single-degree-of-freedom (SDOF) square cylinders in staggered arrangement were investigated to guide the design, processing and manufacture of heat transfer components. Numerical simulation on the FIV of two square cylinders in staggered arrangement was carried out at Reynolds number Re = 100 with the mass ratio of 20. The wake structures of two square cylinders in different arrangements were analyzed. Three distinct flow regimes, i.e. one single-street modes S-I and two double-street modes T-I and T-II, are identified based on Strouhal numbers and flow structures. The amplitude and frequency of FIV response and the lift coefficient of two square cylinders at Ur = 1–30 were obtained to analyze the FIV behavior. The results show that upstream cylinder in the arrangements of mode S-Ia① and S-Ia② is prone to galloping while downstream cylinder in the arrangements of mode S-Ia③ and T-I is prone to galloping. The VIV amplitude of downstream cylinder in most staggered arrangements is much larger than that of single square cylinder while the VIV of upstream cylinder is suppressed.
Study on the Influence of Roughness on Fuel Rod Gap Heat Transfer
Authors: Liu Weichao , Zhang Guoliang , Li Changzheng , Cao Jin , Lin Shaofang , Fu Xiangang , Nie Lihong , Deng Yongjun
Published: November 2022

The behavior of pellet-cladding gap is very complex, which includes many physical phenomena, such as the change of gap size caused by pellet swelling and cladding creep, the change of gap gas composition caused by fission gas release, and the change of thermal conductivity of single component and multi-component gas.

Pellet-cladding gap heat transfer coefficient is a key parameter for safety analysis, which directly affects the thermal-mechanical performance of fuel rods under different conditions. Among them, pellet-cladding gap is the key parameter affecting pellet-cladding gap heat transfer. Because the roughness of the outer surface of the pellet and the inner surface of the cladding directly affects the size of the gap, it is necessary to conduct in-depth research on it.

The research on the influence of pellet and cladding roughness on pellet-cladding gap heat transfer focuses on the experimental study and model building, but seldom on the influence of the irradiation process in the fuel reactor.

On the basis of studying the steady-state and transient operation of fuel rods and the influence of roughness on various gap heat transfer modes, the influence law and mechanism of roughness on pellet-cladding gap heat transfer are discussed, which has guiding significance for the design of new fuel rods.
A Digital Proof of Concept (POC) for Simulating the Coupled Phenomena Between Neutronics, Structural and Fluid Dynamics in a Reactor Core
Authors: Liu Jie , Zhang Chunyu , Zheng Zhihao , Min Jiesheng
Published: November 2022

Numerical simulation is a key technology for improvement of nuclear reactor design and maintenance with its low cost and safety comparing to traditional experiments, especially when irradiation is involved. Various software exist today in the fields of structural mechanics fluid dynamics and even neutronics to model and understand these phenomena. Nevertheless, coupled simulation of the reactor core represents a span of different mechanisms with intricate links between neutronics, materials and the surrounding coolant. With increasing computing power, advanced coupling of the phenomena occurring in the reactor core is being developed nowadays.

In the present study, a multi-physics coupling framework is developed by coupling three codes, i.e., a neutronics transport code, a computational fluid dynamics code and the finite element structural analysis code. The three solvers are hosted within the open-source platform SALOME. Besides this framework, the platform also contains additional modules to realize the full process of a multi-physics simulation. The multi-physics coupling framework provides a strong scalability, users are able to add user-defined modules though open API interface.

This paper provides a route forward for the development of the framework, along with the description of its various anticipated modules and functionalities. Ultimately, it is expected that the multi-physics coupling framework will achieve the full process of a multi-physics simulation, including preprocessing, neutronics calculation, structural analysis, CFD (Computational Fluid Dynamics) simulation, multi-physics coupling simulation, post-processing, and data analysis.
The Study of Removal of Secondary Neutron Source of HPR-1000
Authors: Xiao Huiwen , Li Xiang , Liu Guoming , Zhang Yicheng , Shao Zeng , Xu Min , Yi Xuan , Yang Haifeng
Published: November 2022

When the reactor was in the subcritical state, the criticality safety of the core was monitored by the source range detector which was placed out of the core. According to the “Initial Loading Test for Pressurized Water Reactor of Nuclear Power Plant” (NB/T 20434-2017RK), the recommended value of count rate of the detector was larger than 0.5 cps. But during loading and other subcritical reactor conditions, the number of neutrons in the core is relatively low, causing that the count rate of the ex-core source range detectors could not reach 0.5 cps. Thus, secondary neutron source was placed in the following cycle of reactor to increase the neutron count rate of the ex-core source range detectors to 0.5 cps to meet the standard. In fact, the use of secondary neutron sources raises a number of problems, such as: 1) the production of radioactive and other tritium from secondary neutron sources, the total production of which is a limiting criterion for the construction of nuclear power plants; 2) the increased procurement costs of secondary neutron sources; and 3) the vulnerability to damage due to the long presence of secondary neutron sources in the core, increase the radioactivity level of the primary circuit. Because of the use of low leakage loading fuel management in HPR-1000, the spent fuel assemblies were placed closer to the source range detector and generated more neutrons, it is possible for the ex-core source range detectors to meet the counting requirement of 0.5 cps regardless of whether the core is in the subcritical state. This paper studied the removal of secondary neutron source in HPR-1000. The neutron source intensity of spent fuel assemblies and neutron count rate in fuel loading process were calculated in a conservative method with three dimension monte carlo code. The calculated results had shown that the removal of secondary neutron source in HPR-1000 satisfies safety requirements.
Application Research on Human Reliability Analysis Method of Class C Human Action of China Fast Reactor
Authors: Niu Jingjuan
Published: November 2022

According to the current situation of human reliability analysis of China fast reactor, combined with the operation experience of China Experimental Fast Reactor (CEFR), this study not only identifies the method of important personnel actions, but also standardizes the technical process of human reliability analysis. Compared methods of four kinds human reliability analysis, finally, the SPAR-H method is choose for CLASS C. According to the mathematical model of this methods, the calculation code for system analysis is developed using Visual Basic language. Based the standardization of the technical process, the code using examples CLASS C are provided, finally the standard technical documentation of HRA is edited. Based on the results of the calculation analysis, some recommendations are provided for systems design, human-machine interface design, regulation design and training program. The above is only an example. In the future, it is necessary to analyze more human actions of CLASS C, so as to give suggestions on the impact of human factors in various fields. To achieve the goal of minimize the impact of CLSS C human actions on the design and application of the whole project.
Commissioning and Cold Test of HLLW Vitrification Plant of China
Authors: Ruan Minzhi , Liu Yingzi , Zhang Jinyao , Sun Minhang , Liu Xuelian
Published: November 2022

Prior to radioactive operations, the Vitrification Plant of China (VPC) will implement a cold-commissioning plan or strategy that will include functional test, water test, simulant test and glass pouring test to not only test the process systems, auxiliary systems and melter performance, but also allow operators to increase facility operating experience and be familiar with remote handling in hot cells before introducing radioactive material into the plant. The process technique used is based on the German liquid fed ceramic-lined electric waste glass melter (LFCM) designed to process high level liquid waste (HLLW) containing highly noble metals (Ru, Rh, Pd).

Several batches of simulants tests are performed and the operation data including melter temperature and initiating procedure are analyzed. A maximum process rate is determined according to the applied process parameters including the high operating temperature. And in different operating modes, including idling mode and operating mode, the effect of different feed rates on the formation of the cold cap and the temperature in different areas of the melter are investigated.

The paper gives an overview about the current status of the VPC plant. The strategy of commissioning and cold test operations are described in detail.
Development of Drift Flux-Type Correlations Considering Bubble Induced Turbulence in Vertical Gas-Liquid Metal Two-Phase Flow
Authors: Wang Di , Liang Ren , Lin Zhikang , Ouyang Yong , Shen Yonggang , Lu Xianghui , Qiu Suizhen
Published: November 2022

In hypothetical steam generator tube rupture (SGTR) accident scenario of pool-type liquid metal cooled fast breeder reactor, water is injected from SG into primary circuit through the break and flash. The injected steam bubbles may transport into the core region and insert positive reactivity, which threaten the reactor safety. To accurately predict the void fraction for evaluation of accident consequence, drift flux model applied to gas-liquid metal two-phase flow is useful and need to be developed. In this contribution, an algebraic model aimed at bubble induced turbulence (BIT) is introduced to improve the underestimation of turbulent viscosity calculated by Clark’s force balance model to accurately predict the one-dimensional liquid velocity distribution along radial direction in flow channel. On this basis, the distribution parameter is calculated. Analysis result shows that distribution parameter is assuming very high values at low Re number. With Froude number increase, distribution parameter tends to decrease. At lower void factions, distribution parameter is also assumed to be high values. It indicates that the pipe size, flow rate and void fraction can all influence distribution parameter. Considering the quantitative laws of above influence factors obtained by theoretical analysis and fitting test data, a new correlation for gas-liquid metal two-phase flow is developed and evaluated.
Research on the Supercharging Performance of the Annular Multi-Nozzles Ejector at the Suction Port of a Centrifugal Pump
Authors: Ma Fubing , Fan Guangming , Xu Junxiu , Yan Changqi , Hao Shuai , Zhou Yuning
Published: November 2022

The coolant may vaporize in the centrifugal pump to cause cavitation in pool-type nuclear heating reactor. Installing a jetting device at the pump suction port is an effective means of applying jetting technology to solve the problem of preventing cavitation in centrifugal pumps. In this paper, an annular multi-nozzles ejector was designed and an experimental system was constructed to test the actual supercharging performance of the device. When the high-energy fluid at the pump outlet flowed through the device, it was first mixed uniformly in the annular chamber, and then injected through the nozzles to mix with the main flow to increase the pressure of the main flow. The static pressure at the suction port of the pump increased. Accordingly, the available Net Positive Suction Head (NPSHa) of the pump increased. The test was constructed around the total flow flux through the pump invariable and the main loop flow flux invariable two schemes. The results show that when the proportion of the return flow flux was same, the greater the total flow flux through the pump, the higher the pressure increment at the pump suction port; when the proportion of the return flow flux was larger, the difference in the pressure increment corresponding to different total flow fluxes was larger. When the flow flux through the main circuit of the system was same, the pressure increment increased nonlinearly with the increase of the return flow flux. Increasing the return flow flux was beneficial to the supercharging performance of the jetting device.
Joint Extraction of Entities and Relations Based on Hybrid Feature Representations
Authors: Kun-peng Xing , Yang Xue , De-yan Kong , Wei Dong , Zhen-yan Ji
Published: November 2022

Although the fine-tuning pre-training model technique has obtained tremendous success in the domains of named entity recognition and relation extraction, realistic scenarios exist with many triples of nested entities and overlapping relations. Existing works focus on solving the overlapping triple problem where multiple relational triples in the same sentence share the same entity. In this work, we introduce a joint entity-relation extraction framework based on hybrid feature representation. Our framework consists of five primary parts: constructing hybrid feature representations, bidirectional LSTM encoder, head entity recognition module, entity type classification, and relation tail entity recognition. First, we fuse character-level vector and word-level vector representations via a max-pooling operation to enrich text feature information. Second, the hybrid feature representation is fed into a bidirectional LSTM to capture the correlation between characters and entities. Third, the head entity recognition module employs two identical binary classifiers to detect the start and end positions of entities separately. Then the entity type classification module filters out entities classified as non-entity types by softmax. Finally, we regard relation tail entity recognition as a machine reading comprehension task to eliminate the problem of entity overlap. Specifically, we regard the combination of head entities and relations as queries to query possible tail entities from the text. This framework efficiently handles the polysemy problem, considerably enhances knowledge extraction efficiency, and accurately extracts overlapping triples in domain texts with complicated relationships.
Preliminary Investigations on the Effect of Cooling Water Boundary Conditions on PCS Performance
Authors: Fei Likai , Gao Bin , Fan Pucheng , Yuan Haowei , Wang Husheng , Yang Fuming , Du Lianyue , Zheng Lin
Published: November 2022

The coverage of cooling water film outside the containment shell is one of the most important boundary conditions for Passive Containment Cooling System (PCS) of AP/CAP reactor. The reduced coverage of the water film reduces the heat removal capacity of the containment vessel, which may result in the temperature and pressure within the containment can not be kept at a safe level. In this paper, the effect of cooling water flow rate and water film coverage on the heat removal capacity of PCS was studied experimentally under the loss-of-coolant accident (LOCA). Based on the CAP1400 passive containment water distribution test facility, the cooling water boundary conditions were set according to the combination of different cooling water flow rate and water film coverage rate, and the mass energy was released corresponding to the typical design basis accident condition. The response of parameters within the containment obtained from the test was compared and analyzed with the design condition. The experimental results show that the water film coverage rate has a greater impact on the parameters response within the containment than that of the cooling water flow rate.
Neutronics Design and Modelling of a Nuscale-Based Small Modular PWR
Authors: Qiu Yanfei
Published: November 2022

Small Modular Reactors (SMRs) are characterized by high safety, small size, and modular construction. In this paper, an integrated design of a small nuclear reactor, with thermal power of 200 MW and fuel enrichments below 5%, is designed and analyzed to study the burnup behaviour and reactivity coefficients. The core has been simulated by using the transmission neutron Monte-Carlo program Serpent. The results establish that the core is able to provide an 18-month operating cycle. Comparisons of different enrichments and loading patterns are carried out to explore potential future improvements. The relationship between the refuelling cycle length, geometric size of the core, fuel enrichment, arrangement of the burnable absorber and operating power is demonstrated. It is also confirmed that the SMR design has a sufficient negative feedback coefficient to ensure nuclear safety.
Numerical Investigation on Thermal-Hydraulic Characteristics of Supercritical CO2 in Multiple Compact Channels
Authors: Liang Guohu , Chen Deqi , Jin Feng , Hu Lian , Bu Shanshan , Liu Hanzhou
Published: November 2022

Printed circuit heat exchanger (PCHE) is promising in the application of energy conversion system with its high compactness and efficiency. In this paper, the flow and heat transfer performance of supercritical carbon dioxide (SCO2) PCHE are numerically investigated with multiple channels and different channel lengths. The results show that the heat transfer performance of the Zigzag channel PCHE is increased more than 11% comparing to the wavy channel, while it brings more flow resistance. The pressure drop in the Zigzag cold and hot channels was about two and four times of the wavy channel, respectively. It is recommended for the wavy channel PCHE on account of its better overall performance factor than zigzag channel PCHE. Through the analyses of flow characters among multiple channels, it shows good homogeneity of flow distribution for the flow distribution coefficient in each channel close to 1.0, which is able to effectively avoid uneven flow distribution resulting in high local temperature of the heat exchanger. Furthermore, the overall performance of PCHEs with different core lengths are compared, presenting nearly 10% difference on the thermal efficiency.
Coolability Evaluation of the Debris Bed on Core Catcher in a Sodium-Cooled Fast Reactor With a Whole Vessel Model
Authors: Yamano Hidemasa , Kubo Shigenobu , Sasa Kyohei , Shibata Akihiro , Hourcade Edouard , Dirat Jean François
Published: November 2022

This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The DHRS consists of two circuits of direct reactor auxiliary cooling system and one circuit of reactor vessel auxiliary cooling system. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most fuel accumulates on the core catcher in a short term.
Light Water-Cooled High-Performance Research Reactor Physics Design and Analysis
Authors: Guo Zihao , Wang Guang , Su Genghua , Shi Xiuan
Published: November 2022

Research Reactor can provide a platform for advanced reactors’ fuel and material irradiation test, medical isotopes production and neutron science research, which plays an important role in nuclear energy and technology domain. Nowadays, most of research reactor in worldwide are reaching the retirement and decontamination stage. For the increasing demand of the material irradiation and medical isotopes in the world, several new and advanced research reactors should be constructed to replace the void in the future. Differed from the design scheme of the traditional PWR, that of research reactor focus on increasing the power density as high as possible for the huge amount of neutron flux. According to the full reference of the existing advanced research reactor, this article starts a reactor physics design scheme and illustrate a unique reactor physics design. The neutron flux, length cycle and instrument flexibility of this design are better than the reference design.
Construction of the Failed Fuel Information Database Platform
Authors: Liu Xin , Li Changzheng , Wang Yang , Zhang Hongyang , Deng Yongjun
Published: November 2022

The fuel failures occur in the process of design, manufacture and operation the whole time, which will affect the reliability of the fuel. In order to study the method of fuel reliability analysis, it is necessary to establish a database platform of failed fuel information. Failed Fuel Information Database Platform (FAFID) platform established for the first in CGN with demand-based peer group. The platform adopts Browser/Server architecture, Tomcat for web application server software, MySQL for database server software and IE browser for client. After the completion of the platform, it realizes the function of saving the key parameters of fuel assembly/fuel rod in the process of design, manufacturing and operation, and the information can also be classified. The results of poolside inspection and hot cell examination are also arranged in the plate model block. In addition, the platform can also compare the differences in key fuel data, intuitively display the current situation of failed fuel in the form of various statistical charts and tables, so as to help technicians analyze the causes of failed fuel fast and efficient. At the same time, it can also help business personnel better predict and analyze the possible sources of failure. The platform has become a tool for business personnel to comprehensively analyze the causes of failed fuel.
Validation of TULIP via ICSBEP Critical Benchmark
Authors: Chen Wenjie , Du Xianan , Zheng Youqi , Wang Yongping , Wang Rong , Wu Hongchun
Published: November 2022

NECP-SARAX is a neutronics code system for fast spectrum reactor developed by Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In previous work, NECP-SARAX has shown high performance on fast spectrum reactor analysis. Recently, neutron-moderating materials are employed in advance reactors design where the pure fast spectrum is softened to intermediate and thermal energy spectrum. Due to the larger fission cross-section below the fast energy range, the volume of reactors reaching criticality can be reduced. Compared that in fast energy range, the temperature reactivity negative feedback resulting from the Doppler effect in thermal spectrum range is more significant, which is conducive to the safety and miniaturization of the reactors. To meet the design requirement of this kind of reactor, the assembly-wise neutron spectrum calculation module TULIP of NECP-SARAX has recently been extended to generate the cross sections for both the thermal and fast spectrum reactor system. Therefore, in this paper, the validation works of TULIP code have been performed. In order to systematically validate the accuracy of TULIP code, a series of benchmarks with neutron-moderating material are selected from the ICSBEP, such as HEU-MET-FAST-001-002,HEU-MET-FAST-027-001, U233-SOL-THERM-015-001. The numerical results showed that the TULIP code had accurate neutron spectrum calculation capability for the advanced nuclear reactor design.
Feasibility Study on Power Ramp Test Under Atmospheric Pressure and Ordinary Temperature
Authors: Wei Xiangyu , Zhang Wenhua , Zhao Yingchun
Published: November 2022

Pellet-to-cladding mechanical interaction is an important physical phenomenon during reactor power change as well as a multi-phenomenal fuel rod failure mechanism involving stress, strain and material irradiation properties. In order to avoid the failure caused by PCMI, a large number of power ramp tests have been carried out by international organizations over the past decades. The typical way of power ramp test in a high temperature and pressure loop requires stringent test capabilities and high test costs. The key parameters of the PCMI phenomena are stress and strain of cladding, which are generally chosen as the evaluation indicators of PCMI. If it is possible to simulate the pellet-to-cladding contact state, i.e. the gap between pellet and cladding under basic irradiation power level in atmospheric pressure and ordinary temperature environment, then the high stress and strain state of the cladding during the subsequent power ramp test could also be simulated in the same environment, which means lower test costs and test loop requirements. Therefore, a sensitivity analysis using the fuel rod performance analysis code RoPE, was carried out on factors such as initial pellet-to-cladding gap and fuel densification in the power ramp test rod design. By adapting the manufacturing parameters of the test rod and coolant conditions, the high stress and strain state of the cladding could be simulated in the test environment at normal temperature and pressure. The sensitivity analysis provides a theoretical basis for conducting power ramp tests in an atmospheric pressure and ordinary temperature loop.