bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Technical Scheme and Frame of Extreme Hazards Mitigation Guideline for CPR1000 NPP
Authors: Zhang Juanhua , Zhai Shouyang , Liao YeHong
Published: October 2017

The extreme hazards such as large floods, earthquakes, fires or explosion may be lead to an extensive damage or large scale incident in NPP, for instance, loss of command and control, loss of most instruments or equipment which are used for accident mitigation. The typical situations are loss of control room and alternate shutdown capabilities, or loss of AC and DC power, or all of them. Extreme Hazards Mitigation Guideline (EHMG) is developed as the coping strategy for the above extensive damage situations according to the requirements of the characteristic of CPR1000 NPP and the modifications implemented after Fukushima accident. The technical scheme and frame of EHMG includes two modules: EH-IRs (initial response) and EH-MGs (mitigation guideline). And then based on the typical accident analysis, the critical diagnostic parameters and mitigation strategies were made for the EHMG. EH-IRs contains the following contents: Rebuilding the command and control, Off-Site and On-Site Communications, Initial Operational Response Actions, Initial Damage Assessment. EH-MGs contains the following strategies: Manually operating auxiliary turbo pump, Manually depressurizing SGs and injecting into SG, Alternate makeup to RCS, Controlling containment conditions, Alternate Makeup to Spent fuel pool, Alternate Makeup to water storage tank, Containment flooding and so on. The strategies in EHMG can cover both prevention and mitigation phases of severe accident. The first EHMG in China has been validated and implemented in Hong Yan He NPP in September, 2016.
DEM-CFD Study of Purge Gas Flow in Pebble Bed of WCCB for CFETR
Authors: Chen Youhua , Chen Lei , Liu Songlin
Published: October 2017

As one of the breeding blanket candidates for China Fusion Engineering Test Reactor (CFETR), the water-cooled ceramic blanket (WCCB) was proposed to use mono-sized beryllium pebble bed and binary mixed Li2TiO3/Be12Ti pebble bed in order to increase the packing factor and meet the tritium breeding ratio requirement. Helium (mixed with 0.1% content of H2) is used as the purge gas to sweep tritium out when it flows through the pebble beds. Purge gas flow characteristics are of great importance to tritium recovery system design and will dominate the tritium sweep capability. In this study, DEM-CFD method was used to study the flow characteristics including porosity distribution, velocity distribution and pressure drop in the pebble beds. Mono-sized pebble bed with a packing factor of 0.61 and binary mixed pebble beds with the diameter ratio of 6:1, packing factor of 0.755, and diameter ratio of 7:1, packing factor of 0.7513 were simulated. This method can be used to study the detailed flow characteristics in pebble beds and optimize the pebble bed packing parameters to obtain an appropriate pressure drop, and could be extended to study tritium sweep capability for the design of fusion blanket.
Analysis of Hydrodynamic Characteristic in Straight-Tube Sodium-Heated Once-Through Steam Generator
Authors: Zhong Xianping , Yu Jiyang , Saeed Muhammad , Yang Chunzhen , Zhang Xiaolong
Published: October 2017

This paper presents a thermal-hydraulic model of the straight-tube Once-Through Steam Generator (OTSG) of the Sodium-cooled Fast Reactor (SFR) and its application to analyze the hydrodynamic characteristic in an OTSG. The calculated region is divided into four regions, namely subcooled region, nucleate boiling region, film boiling region, and superheated region. Drift-flux model is adopted for two-phase region. The governing equations are based on the mass, energy and momentum equations. The hydrodynamic characteristic is obtained from numerical iterations of the steady-state governing equations. A code named SGCODE is developed by this model and is verified with nominal design data of this OTSG. The influence factors including inlet water throttling coefficient, inlet sodium volume flow rate, and inlet sodium temperature are analyzed.
Numerical Investigation on Conjugate Heat Transfer of Supercritical CO2 in a Horizontal Double-Pipe Heat Exchanger
Authors: Zhao Zhenxing , Wu Jun , Lin Yuansheng , Xiao Qi , Bai Fan , Liu Zhouyang , Liu Yong , Tao Mo , Dai Chunhui
Published: October 2017

The special fluid flow and heat transfer characteristics of supercritical CO2 in a horizontal double-pipe heat exchanger have been numerically investigated. The AKN k-epsilon model was selected to model the turbulent flow and heat transfer of supercritical fluid. In conjugate heat transfer process, there exists obvious heat transfer deterioration on the top wall for horizontal flow. The region of heat transfer deterioration expands with the increased GShell or TShell,0, and the influence of TShell,0 on conjugate heat transfer is greater than that of GShell. The high-temperature fluid will gather near the top region. The intensity and position of the secondary flow can represent the turbulence heat transfer. When the supercritical fluid temperature is much higher than Tpc, buoyancy force can be omitted, but it can not been neglected even under relatively high mass flux.
Effect of Mesh Resolution and Buoyancy-Driven Flows on the k–ε Turbulence Models Using the HYDRAGON Code
Authors: Saeed Muhammad , Yu Jiyang , Xianping Zhong , Hou Bingxu , Zhang Xiaolong , Ghazanfar Mahmood Ahmad , Guangyin Liu , Yang Chunzhen
Published: October 2017

Hydrogen may be released by Fuel and Cladding interaction with the steam at very high temperatures into the containment during a severe accident at the nuclear power plant (NPP). Locally, high hydrogen concentration may be achieved that might probably result in detonation or fast deflagration and challenges the reliability of the containment. The mixing and distribution of hydrogen is a serious safety issue for scientists to preserve the structural reliability of the containment. Although the Three Miles Island (TMI) accident in 1979 was the initiator to study the production and buildup of hydrogen in the containment. Subsequently, the hydrogen explosion in the Fukushima Dai-ichi NPP accident (2011), modeling of the gas behavior turn out to be a significant topic in the nuclear safety analysis.

Computational fluid dynamics (CFD) codes can be used to investigate the hydrogen distribution in the containment during accidental scenarios and predicts the local hydrogen concentration in different regions of the containment vessel. In such a way, the associated risk of hydrogen safety can be determined, and safety assessment and procedures can be measured. The current paper presents the results of systematic work done by using the HYDRAGON code, developed by Department of Engineering Physics, Tsinghua University, to study (I) Mesh sensitivity (II) Buoyancy-driven flows analysis, for various turbulence models i.e., a standard k-ε (SKE) model, a renormalization k-ε (RNG) model and a realizable k-ε (RLZ) model and, to demonstrate the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the NPP. The HYDRAGON code simulation results were compared to the published data performed by Jordan et al., and it has been observed that the simulation results obtained by refined mesh have given generally satisfactory results. However, the global effect of the buoyancy term in the turbulence equations on the HYDRAGON code simulated results is very small.
Numerical Simulation on Steam-Water Separator Performance Based on Population Balance Model
Authors: Xuelong Yang , Guohua Gan , Jing Feng , Qian Zhang , Xuhui Xu
Published: October 2017

The performance of the moisture separator is simulated and analyzed by the Euler two fluid model with population balance model (PBM), and compared with the traditional single drop size calculation method. The traditional Euler two fluid model based on single drop size was used to calculate the drop size effects on the separator performance, and the drop size range of significant effect on the separator was determined. The PBM model was used to calculate the separation performance in the case of multiple drop sizes, and the drop size was divided into five groups with different drop diameters. The percentage of each group was given at the separator inlet. The results show : (1) With increasing drop size, the pressure drop of the separator first increases and then decreases, and the separation efficiency gradually increases and when the drop size is larger than 150 μm, the separation efficiency is close to 100%, therefore, when the PBM model is used, its drop size range can be calculated by single drop size method; (2) With the drop size distribution moving to the direction of large diameter, the resistance coefficient decreases, the separation efficiency increases, the PBM model is consistent with the mass weighting performance from the single drop size method, but the separation efficiency and resistance coefficient are larger. (3) Steady-state PBM model can realize simultaneous calculation with different drop sizes, calculation cost is slightly larger than the single drop size method, but can quickly get the separator performance under a drop size distribution; (4) The unsteady PBM model can automatically obtain a drop size distribution before entering the separator rotating vane, and solve the problem of drop size setting at the inlet of the separator.
Online Fault Diagnosis of Nuclear Power Plants Using Signed Directed Graph and Fuzzy Theory
Authors: Wu Guohua , Zhang Liguo , Tong Jiejuan
Published: October 2017

When nuclear power plant (NPPs) is in fault, it may release radioactivity into the environment. Therefore, extremely high safety standards specification are required during its working. So it is critically important for fault detection and diagnosis (FDD). NPPs are composed of large and complex systems, it is of great significance to obtain the up-to-date information of NPPs’ running state. So FDD is used to provide the state of system accurately and timely in NPPs. Signed directed graph (SDG) can show the complex relationship between parameters and has advantages of conveniently modeling, flexible inference and so on, so SDG is adopted for FDD. To achieve SDG inference better, fuzzy theory is utilized for signal processing in the paper. Firstly, SDG model is built according to the basic steps and principles of SDG modeling, and the parameters are divided into three states which is monitored by fuzzy theory. Secondly, according to the status of parameters, SDG is used for FDD and to reveal the fault propagation path, thus possibility of each fault occurred is achieved. Finally, to verify the validity of the method, the simulation experiments are done for NPPs and the simulation experiments show that SDG-fuzzy theory framework for FDD can get the fault possibility and deeply explain the reasons of fault.
The Research of SFR Aerosol Deposition Behavior in Ventilation Ducts
Authors: Zhao Lei , Ren Lixia , Hu Wenjun , Li Shirui
Published: October 2017

Agglomeration and deposition of particle in a turbulent channel flow are studied. A model about agglomeration and deposition of sodium aerosol particles through the emergency ventilation system is developed. Based on this, the model has been coupled with CONTAIN-LMR. In this way the CONTAIN-LMR could calculate the deposition of sodium aerosol particles in the emergency ventilation pipe. It has been found that the deposition rate is largely determined by gravity, turbulence, inertia and thermophoretic force. The calculation indicates that the deposition rate is about 27%, and it could not be ignored.
Public Perception of Nuclear Power Risk in China and Methods to Improve Public Acceptance
Authors: Shen Jian , Liu Mingye
Published: October 2017

At present, the most influencing factor on nuclear power’s development in China the public’s attitude and acceptance. This paper studies the public perception of nuclear power risk in China, and provides several feasible methods to improve the quality and effectiveness of public perception level. Therefore, the public acceptance of nuclear power can be ameliorated greatly, which will help the development of nuclear power industry.

For decades, the environment pollution has become one of the most serious and urgent problems in China. Since nuclear power has been proved to be a type of low-carbon and environment-friendly energy, striving to develop nuclear power is a good solution to China’s environment issues. However, by the end of 2015, China’s nuclear power’s electricity production share was only 3.03%, which was far below the average level of developed countries. This situation might be partly due to technical and economic reasons, but the essential cause of the restricted development of nuclear power in China is the public perception of nuclear power risk is far from objective and comprehensive, which leads to the public acceptance of nuclear power is not as high as expected.

This paper states that public perception process of nuclear power risk is a dynamic, complex and closed system, which consists of the risk, the transformation of the risk and related information (both truth and rumors), the public perception process and public’s acceptance of the risk, and the public’s actions after receiving the information. The public’s actions often react on the risk perception. Nuclear power risk, unfortunately, is usually magnified, and it makes people become more frightened and oppose nuclear power more seriously.

In order to solve those issues, in this paper, the public risk perception’s characteristics and external influencing factors are studied, a model describes the public perception process of nuclear power risk is developed and analyzed, and the causes of current acceptance level of nuclear power (which is relatively low) are explained. In addition, based on this model, methods to conduct effective risk communication and public education on nuclear power are provided, the future of nuclear power in China can be much better.
An Integrated EDMG to Deal With Extensive Damage for NPPs in China
Authors: Yun Yu , Bo Zhao , Xinli Yu
Published: October 2017

Fukushima Daiichi accident shows that extreme external events may lead to extensive damage in nuclear power plants (NPPs), which may not only affects equipment, but also damage the control and monitor ability of control rooms (including human factor, habitability, control panel and so on). Accidents-related procedures of NPPs in China are Emergency Operation Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs), both of them rely on control and monitor ability of control rooms. Therefore, current structure of accident management procedures can’t work effectively in the scene of extensive damage. The development of Extensive Damage Mitigation Guidance (EDMGs) may strengthen the defense in depth. In relation to this, an Integrated EDMG has been developed to deal with extensive damage resulted from fire (or explosion) or extreme hazards, while the Nuclear Energy Institute (NEI) EDMG only considers fire and explosion. Besides the extended scope, Integrated EDMG develops a comprehensive and detailed mitigation method which is applicable for extensive damage. In this paper, the characteristics of extensive damage will be analyzed first, and then the structure of Integrated EDMG is developed based on the characteristic analysis. According to the emergency response analysis and accident management research, detailed introduction of the Integrated EDMG and its sub-guidances are discussed.
Effects of Neutron Irradiation on the Mechanical and Electromagnetic Properties of Reactor Pressure Vessel Steels
Authors: Chengliang Li , Guogang Shu , Jun Chen , Yi Liu , Wei Liu , Yichuan Hu , Yuangang Duan , Feihua Liu
Published: October 2017

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.
An Extended Application Research of RELAP5 to Typical Non-LOCA of Passive NPP
Authors: Haitao Wang , Zhan Liu
Published: October 2017

As to passive nuclear power plant (NPP) transient/accident analysis, conservative codes rather than best estimated code such as RELAP5 are mainly used to do the non-LOCA analysis work. While considering the severer accident consequence, RELAP5 is selected in some cases as a tool for LOCA analysis so that safety margin can be gained and meanwhile, the safety NPP condition is guaranteed. In order to extend the RELAP5 application rang to passive NPP, a simulation capability research of RELAP5 to typical non-LOCA consisting of condition II, III and IV transient analysis of CAP1000 was conducted, and the results were compared with that analyzed by safety analysis code. Meanwhile, as to simulation skill, an innovative RELAP5 model was developed. In despite of different conservative degree of the key parameters, the response of the key equipment together with the variation tendency of the key parameters were consistent with that predicted by safety analysis code. Besides, the results met the acceptance criteria. It was showed that RELAP5 has the capability to simulate the CAP1000 non-LOCA.
Status of MELCOR Sodium Models Development
Authors: Louie David L. Y. , Humphries Larry L.
Published: October 2017

A sodium coolant accident analysis code is necessary to provide regulators with a means of performing confirmatory analyses for future sodium reactor licensing submissions. MELCOR and CONTAIN, which have been employed by the U.S. Nuclear Regulatory Commission for light water reactor licensing, have been traditionally used for Level 2 and Level 3 probabilistic analyses as well as containment design basis accident analysis. To meet future regulatory needs, new models are being added to the MELCOR code for simulation of sodium reactor designs by integrating the existing models developed for separate effects codes into the MELCOR architecture. Sodium properties and equations of state, such as from the SAS4A code, have previously been implemented into MELCOR to replace the water properties and equation of state. Additional specific sodium-related models to address design basis accidents are now being implemented into MELCOR from CONTAIN-LMR. Although the codes are very different in the code architecture, the feasibility fit is being investigated, and the models for the sodium spray fire and the sodium pool fire have been integrated into MELCOR. A new package called Sodium Chemistry (NAC) has been added to MELCOR to handle all sodium related chemistry models for sodium reactor safety applications. Although MELCOR code requires the ambient condition to be above the freezing point of the coolant (e.g., sodium or water), the high relative freezing point of sodium requires MELCOR to handle situations, particularly far from the primary circuit, where the ambient temperatures are usually at room temperature. Because only a single coolant can be modeled in a problem at a time, any presence of water in the problem would be treated as a trace material, an aerosol, in MELCOR. This paper addresses and describe the integration of the sodium models from CONTAIN-LMR, and the testing of the sodium chemistry models in the NAC package of MELCOR that handles sodium type reactor accidents, using available sodium experiments on spray fire and pool fire. In addition, we describe the anticipated sodium models to be completed in this year, such as the atmospheric chemistry model and sodium-concrete interaction model. Code-to-code comparison between MELCOR and CONTAIN-LMR results, in addition to the experiment code validations, will be demonstrated in this year.
Analyzing Fuel Bundle Deformation Using the Finite Element Method
Authors: Soni Rabia , Chan Paul K. , Wowk Diane , Williams Anthony
Published: October 2017

During off-normal conditions, the temperature transient experienced by a CANDU fuel bundle may lead to deformation of the bundle via fuel element bowing between the spacer pads due to sub-channel temperature gradients, or fuel element sagging under gravity. The resulting deformation could impact the coolant flow distribution through the bundle, causing further degradation of the fuel cooling. Moreover, the deformation of the fuel elements may also lead to contact with the pressure tube, leading to localized heating that may compromise the integrity of the pressure tube. The objective of this work is to create a 3D finite element model to analyze bundle deformation behaviour under transient conditions. The methodology and preliminary results are presented herein.
Numerical Simulation of Crud Effects on the Flow and Heat Transfer Performance in PWR 17×17 Rod Bundles
Authors: Ren Bing , Gan Fujun , Dang Yu , Zhu Libing
Published: October 2017

Corrosion products on fuel cladding surface have a significant impact on reactor operation. These types of deposits are defined as Corrosion Residual Unidentified Deposit (CRUD) and are consist of a porous matrix of nickel and iron based oxides deposited on the fuel cladding surface. It is well known that crud deposits may cause potential Crud Induced Localized Corrosion (CILC) risk and Crud Induced Power Shift (CIPS) risk.

The paper presents a Computational Fluid Dynamic (CFD) method of predicting the crud effect on the thermal hydraulic performance. The effect of the crud roughness is mainly considered in the simulation, the flow near the wall of the crud is solved by modifying wall function in the prism layer.

The simulation object is a span of typical 17×17 rod bundle with a mid grid in PWR, all the structures including grid straps, springs, dimples, mixing vanes and welding spots are included. Thicknesses of grid and fuel cladding are considered in order to precisely simulate the fluid-solid conjugate heat transfer. The crud is set to be covered on the full span downstream of the grid. The simulation is based on the CILC risk pre-analysis and the computed information in the mostly likely crud deposit position is used as boundary condition.

Based on the simulation results, the crud effects on the flow characteristics including vortex structures, circulation, turbulent intensity and second flow intensity and the heat transfer characteristics including rod temperature, fluid temperature and heat transfer coefficient are discussed in detail.
Comparison Research on Different Aerosol Pool Scrubbing Models
Authors: Gao Shengqin , Fu Yaru , Sun Dawei , Mei Qiliang , Pan Nan , Zhang Shanshan , Huang Gaofeng
Published: October 2017

In order to recommend a pool scrubbing model to be applied in investigating the aerosol pool scrubbing effect of spent fuel pool (SFP) during the period of CAP1400 containment overpressure release, different pool scrubbing models are studied in this paper. At first, the thermal hydraulic processes and aerosol retention mechanisms considered in five mainstream pool scrubbing models are compared briefly, through which SPARC-B/98 model (the updated version of SPARC-90) is identified as the most comprehensive one. Subsequently, ACE aerosol pool scrubbing experiments are simulated using SPARC-B/98 and SPARC-90 model respectively, whose results could contribute to assess the conservatism of the two models. At last, considering the lack of experimental decontamination factor (DF) data under CAP1400 SFP pool scrubbing boundary conditions, sensitivity study involving these special conditions is performed with SPARC-B/98 and SPARC-90 model, in which injected gas flow rate, steam volume fraction in injected gas and pool temperature are selected as the analyzed parameters. The results show that SPARC-B/98 model is more conservative and applicable under SFP pool scrubbing boundary conditions and are advised to be used in analyzing the SFP aerosol pool scrubbing effect.
Thermal Performance Investigation of Hexagonal Spent Fuel Dry Storage Facilities
Authors: Ren Bing , Ni Chenxiao , Dang Yu , Liu Jiazheng
Published: October 2017

A new type of dry storage system is designed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI), which can efficiently remove the decay heat of the hexagonal spent fuel assemblies such as VVER fuel assemblies. The dry storage system includes a Ventilated Concrete Cask (VCC) and a Multi-assembly sealed basket (MSB). Decay heat is removed by natural circulation with helium and air, heat conduction and thermal radiation heat transfer.

Thermal performance of the dry storage system has been investigated by two different numerically methods, i.e., the Computational Fluid Dynamics (CFD) method and the lumped parameter method. The CFD method is utilized based on the commercial software STAR-CCM+, and fuel assemblies are modeled as a porous medium characterized by effective conductivity and the permeability and inertial resistance factor, while other geometry including the lids, base plates, inner and outer shell are modeled explicitly with necessary simplifications. The lumped parameter method is utilized based on the system code GOTHIC, the geometry and the fuel assemblies are divided and represented by 44 volumes. The flow of the air and helium are modeled by flow path which connects the related volumes, and the heat transfer between fluid and solid structures are modeled by thermal conductor models. Heat transfer by convection, conduction and thermal radiation is modeled in both of the two methods.

The maximum temperature of spent fuel assembly can be obtained by both of the two methods, which can be a design basis for investigations attempting to improve the performance of the dry storage system. It is found that the simulation results calculated by the lumped parameter method are more conservative than those calculated by the CFD method. Both methods indicate that after the storage of 7.5 years, the dry storage system is able to remove the decay heat from the hexagonal spent fuel assemblies, keeping maximum cladding temperature below the design limit. Besides, detailed flow characteristic are obtained by CFD simulation. Furthermore, effects of MSB normal operating pressure and the ambient temperature are studied.
CFD Analysis of the Primary Side in a Helical-Coil Once-Through Steam Generator
Authors: Ye Kai , Zhang Yaoli , Lin Jianshu , Li Ning , Yang Yinglin , Li Zhuocheng , Hao Junwei , Chen Yifen
Published: October 2017

The helical-coil once-through steam generator (OTSG) is usually used in the nuclear power plant when the compactness of equipment was taken into consideration. The investigation of flow parameters in the primary side is valuable for the optimization of the OTSG. The purpose of this research is to obtain a further understanding of fluid behaviors in the primary side of the OTSG to achieve a more rational design. Using ANSYS ICEM and ANSYS FLUENT, a three-dimensional (3D) computational fluid dynamics (CFD) model was created and analyzed. Through a series of cases, the velocity profiles and pressure drop through the primary side of the helical-coil OTSG have been calculated, and the influences of different structure designs on the coolant flow parameters have also been tested. Ultimately some pertinent suggestions for improvements were proposed, and insight is obtained into the importance of various modeling considerations in such a model with a complicated structure and large-scale grids.
Development and Verification of Multi-Group Cross Section Process Code TXMAT for Fast Reactor RBEC-M Analysis
Authors: Qiu Ruomeng , Ma Xubo , Xu Qian , Liu Jiayi , Chen Yixue
Published: October 2017

In order to generate cross sections for fast reactor calculation, a code named TXMAT based on object–oriented programming and allocate memory technology has been developed. It has the capability to generate macroscopic cross sections for transport or diffusion calculation and microscopic cross sections for burnup calculation, and can also deal with the ultrafine group cross sections (more than 2000 groups and P5 Legendre order) which TRANSX 2.15 can’t do. It works together with a generalized cross section data library called MATXS to give the transport code users easier access to appropriate nuclear data and capabilities which are difficult or impossible to get with any other systems. The TXMAT can handle the shielding effects of many isotopes through background cross section iteration. Several critical benchmarks are calculated. It is shown that the total cross section, absorption cross section, fission neutron spectrum and zero Legendre scattering matrix have been verified using TRANSX 2.15, and the maximum relative difference for the main groups is less than 0.2%. After the critical benchmark calculation, the RBEC-M benchmark is used for the whole core calculation. It is shown that the effective multiplication factor of the calculation is consistent with that of other codes, and the power distribution is also in good agreement with that of other codes except for the blankets. The maximum relative difference of the power distribution among core-1, core-2 and core-3 regions is less than 2.3%. But in the blankets the relative error is about 33%, which may be caused by the difference of the weight function between IWT = 8 and real model. Further analysis will be performed in the future.
Application of Underwater Electric Discharge Machining in Nuclear Power Plant
Authors: Huang Xiangming , Huang Ran
Published: October 2017

In the process of operation and maintenance of nuclear power plants, it is required to have technology and equipment for cutting, forming and repairing under specific environments such as radiation, underwater and remote control. Traditional machining methods have limitations in this application environment. Electrical Discharge Machining (EDM) is a special machining method that uses a discharge electrode with a specific geometrical shape to conduct electrical erosion on a metal (conductive) part and erode the conductive material, so as to achieve the purpose of processing. In this paper, underwater EDM technology and equipment of nuclear power plant are studied, a kind of underwater EDM device is designed, the working principle of the device is expounded, the structure and system design of the mechanical and electrical system are introduced. At the same time, the application of underwater EDM technology in nuclear power plant was studied in combination with the cutting requirements of welding irradiation monitoring sample of the reactor pressure vessel. Based on the key technologies that have been solved in the research of underwater EDM technology and equipment of nuclear power plant, the overall technical scheme of welding irradiation monitoring sample cutting device is given, and the cutting results are analyzed.
Radioisotope Thermophotovoltaic Generator Design and Performance Estimates for Terrestrial Applications
Authors: Wang Xiawa , Chan Walker , Stelmakh Veronika , Fisher Peter
Published: October 2017

This work provides the design methods and performance estimates of the radioisotope thermophotovoltaic system (RTPV) for terrestrial applications. The modeling is based on an experimentally tested prototype using two-dimensional high temperature photonic crystal to realize spectral control. The design efforts focus on the optimization of the system efficiency and contain the heat source number, the size of the energy conversion elements, the insulation configuration, and the heat sink design. An equivalent circuit model was developed for the thermal and electrical performances. Based on a specific output requirement, an optimized heat source number and energy conversion area can be computed for a certain cell type and insulation design. The selection and characterization of the low bandgap thermophotovoltaic (TPV) cells applicable to the generator are compared and discussed. The generator’s heat sink design uses extended fins and the performance is estimated based on the external operating conditions. Finally, the work provides a design example of a terrestrial RTPV generator with an output level of ∼40 W electrical power (We) using InGaAsSb cell, reaching an efficiency of 8.26%.
Study on Pool Boiling Behavior of the Spent Fuel Under Accident Conditions
Authors: Gao Shang , Lu Daogang , Wang Han , Zhong Yuhang
Published: October 2017

Under normal refueling and emergency full-core offload condition, the fuel assembly is removed to the Spent Fuel Pool (SFP). Decay heat produced by the spent fuel is carried out by the cooling system. Active cooling method is adopted by the traditional PWR nuclear power plants, which means decay heat is taken away depending on forced circulation of the pump. However, the spent fuel pool, under accident condition, will lost the forced circulating cooling capacity, which will be a threat of for the fuel building safety. To study the thermal-hydraulic characteristics in the SFP missing the forced cooling, through CFX methodology and experiment, change of temperature and heat transfer coefficient of the wall of the heating tube at different heights were discussed, meanwhile the streamline chart and temperature contour were obtained as well. The present result indicated that under different power conditions, different height of water temperature increased at first and then trend to stable at saturation temperature. For a single 9*9 spent fuel assembly, water temperature at the higher height is higher than the lower at the same time, and water temperature at higher location reached a stable value more quickly. In addition, power value had a significant impact on the time of reaching saturation temperature, for example, 7000s is needed to reach saturation under 8.68KW condition while only 3000s under 16.12KW, which illustrates that fuel unload power is crucial to the SFP safety. Based on the experiment data and single phase calculation, heat transfer coefficient at different height of the heating tube decreased slowly at first, and then increased. Especially, heat transfer coefficient at the highest test point rapidly decreased at one point because of boiling crisis.
Test Reactor Irradiation Fuel Assembly Development in China
Authors: Ding Jie , Zheng Yixiong , Ding Yang , Liu Song , Zhu Libing , Liu Jiazheng
Published: October 2017

During the development of zirconium alloys, the irradiation in the test reactor is a critical step to comparison the irradiation properties of candidate alloys, such as corrosion, creep and irradiation growth. In this paper, a small scaled fuel assembly for test reactor irradiation is designed, which meets the needs of new zirconium alloys development. The irradiation fuel assembly (IFA) can be easily disassembled, and the test fuel rods or irradiation specimen can be easily replaced, which makes it possible to do the further post-irradiation examination in the hot cell to obtain the irradiation performance data. Now the IFA has finish fabrication and the test reactor irradiation program is planned to launch in 2017.
Research on Estimating Methods and Application of Fuel Rods Defect
Authors: Li Huaibin , Fu Yaru , Mao Lanfang , Mei Qiliang
Published: October 2017

When fuel rods have defects, the fission products in the fuel rods will come out and enter the reactor coolant through the defects. The release of the fission products will result in the increase of radiation dose, the indeterminacy of the fuel management and will influence the economics and safety of the nuclear power station. Based on the analysis of the typical nuclides activities in reactor coolant, the evaluation of the defect fuel rods can be realized. This paper studied the related analysis around the world and determined the methods to evaluate the number, the defect type (open or tight) and the burn-up of the defect fuel rods.

The evaluation method of this paper can be used to evaluate the defects of fuel rods, and can provide valuable information for the fuel management and dose analysis, and also can be a useful technical support to the operation of nuclear power station.
Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)
Authors: Cao Huasong
Published: October 2017

Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB.

A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct.

SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.
Study on Rod Ejection Accident With Coupled Neutronic/Thermal-Hydraulic Code
Authors: Wang Xiong , Li Qiang , Wang Dong , Ouyang Yong , Lu Xianghui
Published: October 2017

The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). REA is usually an asymmetric transient where neutron kinetics and the thermal hydraulics are strongly coupled (through doppler feedback). With the aim of increasing simulation accuracy by eliminating conservatism at code interfaces, the coupled code SMART-FLICA which links 3D-neutron kinetics code SMART and thermal-hydraulic code FLICA III-F is used to perform REA transient. The analysis will characterize the PWR real plant phenomenology more accurately in the RIA category of accidents.
Bubble Diameter Effects on CFD Simulation for Subcooled Boiling Flow in Rolling Circular Tube
Authors: Xu Sheng , Pan Liang-ming , Yang Yu , Yuan Dewen , Xu Jianjun , Huang Yanping
Published: October 2017

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and local parameter distribution with different bubble diameter models in circular tube under static and rolling condition. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be obtained. The calculation results of local void fraction are compared with experiment data to verify the accuracy of the numerical calculation for subcooled boiling flow under rolling condition. The Zeitoun bubble mean diameter model which the most recommended correlation for bubble diameter under low pressure and several fixed bubble diameters are applied to simulate the same condition in low pressure. These results are compared, include the distribution of void fraction, velocity distribution and radial flow induced by rolling motion.

A good agreement with the experimental data has been achieved when Zeition bubble mean diameter and 2 mm fixed bubble diameter are used to describe vapor diameter in static condition. The local void fraction fluctuation has the same period with the rolling motion, and the fluctuation amplitude increases with the increase of rolling amplitude and rolling frequency. The difference shown in rolling condition between calculation results and experimental data demonstrates that better agreement with the experimental data has been achieved in the near-wall region about local void fraction which has bigger fluctuation amplitude. Higher void fraction has gotten using Zeition bubble mean diameter model to describe bubble diameter in subcooled boiling flow, tiny difference has showed in temperature, velocity and radial velocity in different bubble diameter model. Accurate vapor diameter model or method to describe vapor diameter coupled with suitable interphase force model is needed in rolling condition under low pressure to fit the calculation of subcooled boiling better under rolling condition.
Experimental Investigation on Aerosols Collection Performance of Metal Fiber Filter in FCVS
Authors: Tao Ding , Changqi Yan , Xiaxin Cao , Zehua Guo
Published: October 2017

An experimental setup has been designed and fabricated for the analysis of filtration performance of the metal fiber filter as applied to Filtered Containment Venting System (FCVS). The main characteristic of this test facility is the presence of the aerosol and Scanning Mobility Particle Sizer. The objective is to investigate the removal performance of the metal fiber filter for aerosol, as well as further understand the filtration process in the metal fiber filter. It is observed that the metal fiber filter is capable of removing more than 99.955% aerosols at the desired flow rate ranging from 0.17 m/s to 0.3 m/s and the resistance has a significant linear correlation with flow rate. Due to the electrostatic effect, diffusion effect, inertia effect, interception effect and gravity effect, most penetrating particle size plays a significant role in removal performance of the metal fiber filter for aerosol. It is also found that with aerosol size ranging from 0.1 μ m to 0.3μm in most penetrating particle size, the filtration efficiency is more than 99.8% at the flow rate of 0.25 m/s. From this study, valuable reference data and useful information are provided for practical applications.
Preliminary Large Break LOCA and SGTR Accident Analysis of the Compact Small Reactor
Authors: Xing Mian , Li Linsen , Shen Feng , Hu Xiao , Liu Zhan , Qi Zhanfei
Published: October 2017

This paper gives a brief introduction of the Compact Small Reactor (CSR). It is a simplified two-loop reactor with thermal power of 660MW and with compact primary system and passive safety feature. Preliminary safety analysis of the CSR is conducted to evaluate and further optimize the design of passive safety system, especially the passive core cooling system. Large Break Loss Of Coolant Accident (LBLOCA) and Steam Generator Tube Rupture (SGTR) are selected as two reference accidental scenarios. Each scenario is modeled and computed by RELAP5/MOD3.4. For the LBLOCA analysis, a guillotine break happens in the cold leg of the loop containing the core makeup tanks balance lines. The results show certain safety margins from the guideline values, and the passive safety system could supply enough cooling of the core. For the SGTR analysis, the results show the robustness of the design from the safety perspective. It is concluded that the safety systems are capable of mitigating the accidents and protecting the reactor core from severe damage.
Development of Hydrogen Treatment System in Severe Accident: Part 3 — Study of Reactant Diameter Effect on Hydrogen Treatment
Authors: Yamada Akira , Iwaki Chikako , Fujihara Ryoma , Yagyu Motoshige , Okamura Masato , Tanabe Masashi
Published: October 2017

The hydrogen treatment system has been developed in order to prevent the overpressure of the primary containment vessel (PCV) caused by a large amount of hydrogen generated by the metal-water reaction in severe accidents (SAs) of Light Water Reactor. In previous studies, we evaluated the hydrogen treatment rate using a couple of metal oxides, and confirmed that MnO2, CuO, and Co3O4 were effective for the hydrogen oxidation under the oxygen-free condition, then we selected them as reactants[1].

Although the reactants were granulated with a diameter of 2 mm for application to the system, the hydrogen treatment rate has not been scarcely evaluated for the granulated MnO2 which is expected to treat the hydrogen around 120 °C [2].

Thus, we made the diameter of the granulated MnO2 smaller, and found that the hydrogen treatment was occurred by the granulated MnO2 with a diameter below 1.0 mm. The granules with a diameter below 1.0 mm were also acceptable for the system from the point of view of decreasing the differential pressure (DP). Moreover, the experiments using a test section simulating a reactor of the system had been conducted under the hydrogen condition simulating typical condition of a SA, by loading the granulated CuO with a diameter of 2mm onto the granulated MnO2 with a diameter of 1mm. As a result, the hydrogen treatment was markedly accelerated by supplying enough reaction heat from the granulated MnO2 to the granulated CuO.
Preliminary Analysis of Influence of Uncertainty Parameters for CAP1400 LBLOCA Accident
Authors: Qi Zhanfei , Lu Lu , Wang Weiwei
Published: October 2017

The best-estimated methodology is applied for CAP1400 LBLOCA. Based on the sampling and calculation results, the influence of uncertainty parameters including plant initial conditions parameters, core power distribution parameters, global system thermal-hydraulic parameters and local hot rod parameters are evaluated. The major uncertainty parameters are figured out. Moreover, all uncertainty parameters considered in CAP1400 LBLOCA are ranked by their influences. The analysis results show that the break area, relative power of middle third of core, the accumulator initial pressure give the most contribution to PCT of CAP1400 LBLOCA. The uncertainty parameters could be optimized according this paper in the future.
A Hybrid Approach to Seismic and Other External Hazards HRA: More Efficient and More Realistic
Authors: Amico Paul , Collins Erin , Cook Elizabeth
Published: October 2017

As stated in the External Events HRA EPRI guidance [Ref. 1], there are inherent differences from internal events PRA that impact the ability of operators to perform actions during an external event.

“Although the process steps and concepts are the same for an external event PRA, a fire PRA and an internal (non-spatial) PRA, there are several key differences needing to be addressed in the development of an external event HRA. Differences include the impact of the initiating event on the site structures, systems, and components (SSCs). Other differences, such as procedures and the impact on timeline development, should also be considered.”

A damage context “hybrid approach” has been developed and utilized for Seismic Probabilistic Risk Assessment (SPRA) that takes advantage of the binning system developed in the EPRI guidance [Ref. 1], while also applying site and context-specific parameters. In addition, the hybrid approach does not provide overly conservative results, like some scaling and factor based methods. This paper describes the methodology of the Seismic and External Hazards (EH) HRA hybrid approach and provides an example of how it has been applied in a SPRA. This approach meets the Capability Category II quality requirements in Part 5 of the ASME/ANS PRA Standard RA-Sa-2009 [Ref. 2].