bluuebunny/crossref_metadata_2025 huggingface dataset:
The Application of Piping Material Code in the CAP1400 Nuclear Plant Project
Authors: Jie Shen , Zhen-ning Zhang , Yu Liu
Published: October 2017

Nuclear industry differs from most other industries in the characters of large scaling, long period and multiple collaborating institutions [1]. Traditionally, the material codes are compiled respectively by collaborating institutions of nuclear power plant projects, so the whole code system lacks unified management and planning, causing many defects such as the incompleteness in information coverage, the inconformity in classification and description of materials and the confusion in commodity codes. So, it is of great significance in setting up a standard nuclear power material code system to effectively enhance the efficiency of design and management, and to ensure the schedules of the projects.

This article introduces in detail the entire process of the setting up and application of the material code system of State Nuclear Power Technology Company (SNPTC) by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) for the CAP1400 nuclear power plant project. The application of the material code system in the CAP1400 project remarkably simplifies the work of design, material take-off and purchase, improving the project’s quality.
Preliminary Safety Analysis of a Compact Small Reactor: Feedwater Line Break and Small-Break LOCA Accident Analysis
Authors: Li Linsen , Shen Feng , Xing Mian , Liu Zhan , Qi Zhanfei
Published: October 2017

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed.

Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system.

The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.
Periodic Oscillations in Natural Circulation Boiling System
Authors: Chen Xianbing , Gao Puzhen , Wang Qiang , Zhang Yinxing , Liu Jiawei
Published: October 2017

Natural circulation has been widely used in some evolutionary and innovative nuclear power plants. Natural circulation systems are susceptible to flow instabilities which are undesirable in the nuclear power devices. An experimentally investigation of two phase flow instability in up-flow boing channel under natural circulation is presented in this paper. Flow instability with and without flow reversal have been found. A pulse signal of water temperature at the inlet of the test section can be detected when the channel suffers from flow reversal. Single phase and two phase flow alternate in the channel regardless of the occurrence of flow reversal. Periodic oscillations with multiple high-order harmonic waves are confirmed by applying Fast Fourier Transform to the time traces of flow rates. Period of flow instability which is the reciprocal of the frequency with the largest amplitude in the amplitude-frequency plane are obtained. Period of flow oscillation presents a nonlinear change with the increase of mass flux. Period of flow instability increases rapidly with the increase of mass flux and decreases slowly when it reaches the maximum value.
An Experimental Study on Open Natural Circulation Flow Characteristics by Air Injection
Authors: Cao Pengjiu , Cao Xiaxin , Sun Zhongning , Ding Ming , Li Na , He Haisha
Published: October 2017

An open natural circulation system has the characteristics of a simple structure, superior safety performance and strong heat removal capability. However, during long-term operation, the flow instability may occur due to the reduction of the driving force, which will have adverse effects on the heat removal capability and safe operation of the system. Thus, injecting air into the riser is designed in this paper to improve the driving force of the circulation flow, reduce the possibility of flow instability, and increase the heat removal capability. In order to investigate the influence of air injection on the evolution of flow pattern, resistance characteristics and circulation flow rate, the method of visual observation and data analysis is used based on different pore sizes porous media, air injection rate and submergence ratios. The ratio of the driving pressure head to the resistance pressure drop is proposed as the basis for assessing the effect of air injection on the ability of natural circulation. It is found that the driving force of natural circulation increases with the increase of air injection rate, and the circulation flow rate increases obviously when the bubbly flow appears in the riser. However, when the transition from bubbly flow to churn flow appears, the growth of the circulation flow rate slows down because the resistance increases faster than the driving force. Therefore, it can be known that the best performance is obtained when bubbly-churn flow appears in the top of the riser. What’s more, the capacity of lifting water will be reduced and churn flow will appear prematurely when the submergence ratio decreases. This means that in the process of open natural circulation system design, the submergence ratio of the system should be increased as much as possible. Finally, in this paper, it is found that the bubble pump with PS = 0.2 μm has better performance.
CAP1400 Fuel Assembly Development and PWR Fuel Assembly Technology Progress in China
Authors: Zhu Libing , Ding Jie , Zhou Qin , Zhou Yunqing , Gan Fujun , Lu Junqiang , Zheng Yixiong , Li Jiwei , Ding Yang , Zeng Qifeng , Liu Song
Published: October 2017

In 1990s the first generation of PWR fuel assembly FA300 was developed in China. With constantly design improvements over the following 20 years, three types of FA300 fuel assemblies have been developed with max fuel assembly average burnup reach 40000 MWd/tU and successfully used in 300MWe PWRs. By now over 1100 fuel assemblies of FA300 series have been successfully operated in three different 300MWe PWRs with satisfied operation reliability.

The CAP1400 fuel assembly development program was launched in 2010 which includes new zirconium cladding material development, UO2 pellet development, high performance fuel assembly mechanical structure development, fuel rod performance code and fuel assembly seismic analysis code development, out-of-pile fuel assembly test facility construction and in core irradiation program. The main purpose of CAP1400 fuel assembly development program is to meet the needs of self-sufficient fuel supply to CAP1400 reactor.

Based on FA300 and CAP1400 fuel assembly development, a full series of PWR Fuel R&D technology and test system has been successfully established by SNERDI which will continuously support the fuel assembly improving and new type of fuel development. This paper will mainly describe the PWR fuel technology including cladding technology, pellet technology, fuel assembly mechanical design technology, fuel assembly test facility and test technology, fuel rod and fuel assembly code development progress.

Furthermore, the development progress of CAP1400 fuel assembly will also be introduced. By now two types of new zirconium alloys have been selected as candidate alloys, all of the out-of pile performance tests have been finished. Both two alloys show good corrosion resistance. Through full series of fuel assembly components performance tests, the fuel assembly has finished design finalization. Most of the fuel assembly mechanical and hydraulic tests will be finished by the end of 2016. The test reactor irradiation program and commercial PWR irradiation program are also on the schedule. After the LTA program and commercial application licensing, the CAP1400 fuel assembly is anticipated to provide adequate burnup capability and operation reliability to CAP1400 reactor in China.
Thermal-Hydraulic Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment With Modified RELAP5
Authors: Simiao Tang , Jian Song , Dalin Zhang , Chenglong Wang , Suizheng Qiu , Guanghui Su , Wenxi Tian
Published: October 2017

Sodium-cooled fast reactor (SFR) is one of most promising Generation IV reactor technology and has a rapid development in recent years. Experimental Breeder Reactor II (EBR-II) designed by Argonne National Laboratory (ANL) is a typical sodium-cooled fast reactor with a sodium-bonded metallic fuel core, featured with reactor negative reactivity feedback. In order to verify and validate (V&V) the inherent safety performance of sodium-cooled fast reactor, the International Atomic Energy Agency (IAEA) established the coordinated research project (CRP) in which Xi’an Jiaotong University has participated. The “Benchmark Analysis of EBR-II shutdown heat removal tests” was conducted by ANL. SHRT-45R, namely Unprotected Loss of Flow (ULOF), is one of the reactor tests among many transients. Argonne National Laboratory has prepared a detailed benchmark specification and has provided the enough benchmark data for SHRT-45R. At the beginning of SHRT-45R both primary main pumps and the intermediate loop pump were synchronously tripped to simulate the unprotected loss of flow accident. During the test, the plant protection system (PPS) was disabled to initiate a control rod scram. The SHRT-45R test demonstrated that EBR-II could keep in safe during the potentially adverse consequences of unprotected accidents.

This paper introduces the models for predicting SHRT-45R in detail and presents the results of the analysis of the Unprotected Loss of Flow (ULOF) test SHRT-45R performed in the EBR-II reactor. The thermal-hydraulic calculations are performed with Modified RELAP5 in which the thermodynamic and transport properties of liquid and vapor state sodium have been supplemented, as well as the specific heat transfer correlations. The numerical results show that the loss of forced coolant flow causes the coolant temperatures in the instrumented subassemblies XX09 and XX10 to increase to a peak point but keep at an acceptable level about 930K and 850K at the early state of accidents, and then the reactor can shut down by itself due to the negative reactivity feedback including Doppler reactivity feedback and density reactivity feedback after 70s. The variation of the key thermal-hydraulic parameters including the coolant temperatures and fuel cladding temperature in the instrumented subassembly has a good agreement with the experimental data in general. The results could not only verify and validate the inhere safety performance of sodium-cooled fast reactors during potentially unprotected accidents but also demonstrate the analysis capability of the modified RELAP5 for sodium-cooled fast reactors.
Crucial Factors Analysis of 5×5 Grid Flow-Induced Vibration Test
Authors: Zhang C. Z. , Ding J. , Zheng Y. X. , Gan F. J. , Gong S. J. , Mei Y.
Published: October 2017

5×5 small-scale test fuel assembly was manufactured to carry out flow-induced vibration tests, aiming to obtain the vibration responses of new designed grid subjected to axial-flowing water. This test was carried out in a hydraulic loop with the maximum flow rate of 65 m3/h and room temperature and pressure. Laser vibrometer was used to measure the vibrating velocity of grid straps. The vibration peaks were generally distributed in three domains of the frequency spectrum. The first domain is from zero to decades of Hz, in which the vibration may be turbulence-buffeting-induced test assembly vibration as an integer entity. The second domain is from decades to hundreds of Hz, in which the vibration may also be test assembly vibration as an integer entity due to turbopump-induced pressure wave. The third domain is above 1000Hz, in which the vibration may be local strap vibration mainly due to vortex shedding along the edges of the straps. Series of tests were accomplished to figure out the characteristics of the trend of maximum high-frequency vibration velocity and related frequency with increasing flow rate. The high-frequency vibration characteristics of the grid in axial-flowing water are crucial for the fuel assembly designers when evaluating the anti-abrasion performance of fuel rods under reactor flow rate conditions.
Conceptual Core Design of Breeding BWR
Authors: Guo Rui , Yamaji Akifumi
Published: October 2017

High breeding with light water cooling is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. Decreasing the volume ratio of coolant to fuel is normally introduced as a way to harden the neutron spectrum and achieve breeding with light water cooling. Therefore, the tight-lattice assembly was proposed to design reactors cooled by light water with hard neutron spectrum. However, most of them were High Conversion LWRs and none achieved high breeding to meet the growth rate of energy demand in advanced countries.

Tightly packed fuel assembly is designed for the purpose of high breeding. The number ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to less than 0.1 which is about 1/6 of that of Reduced-Moderation Water Reactor (RMWR). Super Fast Breeding Reactor (Super FBR) is one kind of Supercritical Light Water Cooled Reactors (Super LWRs), which adopts these assemblies, obtaining high breeding of CSDT (less than 50 years).

The high breeding performance of Super FBR indicates that, application of the tightly packed fuel assembly on conventional LWR-type reactors, such as BWR-type or PWR-type reactor, may also be effective in achieving high breeding. Compared with Super FBR, the conventional LWR-type reactors with technologies which are currently in use are expected to be easier to implement. When comparing the two main LWR types, BWR-type and PWR-type, BWR-type gains more advantages on breeding, since the coolant is boiling water that generates larger amount of void in the reactor core, leading to a harder neutron spectrum. Meanwhile, from the viewpoint of safety, the negative void reactivity should be satisfied, which is consistent with conventional LWRs. From the viewpoint of neutron economy, high enrichment should be avoided as well.

This study aims to design the BWR-type reactor with the tightly packed fuel assemblies, which attains both high breeding and negative reactivity.
Thermal Hydraulic Performance Analysis for Passive Containment Cooling System of AP1000 in the Event of LOCA
Authors: Bin Mujahid Talha , Yu Yu , Wang Bin , Ali Shahzad Muhammad , Niu Fenglei
Published: October 2017

The design of a nuclear reactor containment building is of key importance in order to enhance the safety of a nuclear power plant. Owing to nuclear accidents such as TMI, Chernobyl and Fukushima, more and more attention is paid to the passive concept in nuclear power development. In order to improve the safety of new generation nuclear power plant, passive systems are widely used, passive containment cooling system in AP1000 is one of the typical example of such kinds of systems. It’s function is to transfer the heat produced in the containment to the atmosphere and keep the pressure in the vessel below the threshold under such accidents as Loss of coolant (LOCA), main steam line break (MSLB), etc. The system operates based on natural circulations inside the steel vessel and in the air baffle outside the containment, and the cooling water is sprayed to the steel surface to enhance the heat transfer process. A proper model simulating the system behavior is needed for system design and safety analysis, and a multivolume lumped parameter approach is employed in order to analyze the containment integrity and to study the long term response of postulated Loss of coolant (LOCA) accidents and Main steam line break (MSLB) accidents. However, the temperature and pressure distributions cannot be described detailed by such model, which is important to study the T-H characteristics in the containment. In this paper LOCA has been simulated on MATLAB using a given pipe break size and the response of containment is analyzed. Furthermore, the results are compared with the results in the Westinghouse Design Control Document 2002. Then the thermal hydraulic performance is studied, the factors such as the air temperature, containment pressure and mass flow rate of the coolant and their effects on the containment are analyzed. This research is done to get further insight on the safety analysis of reactor containment regarding maximum temperature and stress calculation inside the containment.
A Resolution Enhancing Algorithm for Gamma-Ray Spectrum Based on Blind Deconvolution and Lp-Norm Sparsity Constraint
Authors: Li Xinpeng , Fang Sheng , Li Hong
Published: October 2017

Current gamma-ray spectrum analysis method uses a preset system response matrix to improve the resolution of gamma-ray spectrum. However, the system response matrix may not be available or biased due to limitation of experiment conditions, which can degrade the accuracy of gamma-ray spectrum analysis. To solve the problem, a new reconstruction method based on blind deconvolution and sparsity constraint is proposed to improve the resolution of gamma-ray spectrum in this study. The proposed method models the modulation of spectrometer as a convolution operation and reconstructs the high resolution spectrum as well as the convolution kernel simultaneously. Lp-norm based sparsity constraint is imposed to stabilize the demodulation of spectrometer and reduce the background oscillations, so that the resolution can be enhanced. The results of both numerical simulation and experiments demonstrate that the proposed method can effectively improve the resolution of gamma-ray spectrum and reduce background oscillations without any aid of system response matrix.
Research on IVR-Relevant Phenomena of Material Thermodynamic Interaction and Corium Pool Configuration
Authors: Gu Peiwen , Shi Guobao , Cao Kemei , Wang Jiayun
Published: October 2017

In-Vessel Retention (IVR), which arrests relocated molten core materials in the vessel during severe accident, has been singled out as an appealing accident management approach to many reactors. The heat transfer imposed by in-vessel corium is a vital part for IVR success considering the difficulty of significantly altering ex-vessel CHF. For a given decay power, corium pool configuration determines the heat flux profile along the vessel wall, which may produce uncertainties associated with IVR strategy. In this paper, a thermodynamic tool is employed to study the corium pool configurations by analyzing the possible interaction among relocated corium, zircalloy cladding and core internals. The results reveal the immiscibility gap phenomena under high temperature which separates molten materials into oxidic and metal phase in the lower head. The oxidic phase is quite stable and its density is only slightly changed by various accident scenarios. The metal phase is relatively unstable and its density is susceptible to the condition of cladding oxidation degree and crust integrity. The corium pool configurations in the lower head are determined based on the results of thermodynamic analysis and phase density comparison. Both two-layer and three-layer corium pools are likely to be formed under different accident scenarios. CAP1400 has intentionally increased the mass of lower core support plate, which is a beneficial design change to prevent possible focusing effect if material infiltration through crust is assumed to be impossible.
Development of 4S BEPU Method and Application in CAP1400 Containment Response Analysis
Authors: Xin Liu , Benxue Hu , Zhe Wang , Guodong Wang , Zhangli Wang , Di Zhang , Ping Yang
Published: October 2017

Compared with conservation evaluation model, best estimate plus uncertainty (BEPU) method can obtain more realistic results and gain larger license margins with respect to the safety criteria. In view of this, a BEPU method named 4S (SNERDI Statistical Solution for Safety) has been developed, according to the basic principles of evaluation model development and assessment of RG 1.203.

The characteristics of 4S method are as follows: The output uncertainty is quantified by using random sampling and propagation of input uncertainties. Global sensitivity analysis is used to support PIRT establishment. Uncertainties of model parameters are calibrated and validated by using separate effects tests considering measuring uncertainties. DAKOTA code is used for uncertainty and sensitivity analysis. An automatic BEPU analysis platform has been developed by coupling DAKOTA and different reactor safety analysis codes, and code calculations can be performed in parallel.

BEPU analysis of mass and energy release and containment pressure response of CAP1400 under a postulated double-ended cold leg break loss of coolant accident (DECL LOCA) has been carried out by coupling DAKOTA, a mass and energy release analysis code and a containment analysis code. In total, 21 uncertain input parameters are considered. To make the results more stable, the sample size is 124 and the third highest peak pressure is used as the pressure upper bound (with 95%/95% probability/confidence) based on Wilks’ formula. The calculated results show that the peak pressure upper bound is obviously lower than the present conservation method used in license application, with more than 10% analysis margin. Influences of input parameter uncertainties on the containment peak pressure have been analyzed, according to the partial rank correlation coefficients calculated by DAKOTA. The results show that the input parameters mainly affecting the peak pressure are the containment condensation heat transfer multiplier, initial containment temperature, break resistance, decay heat, initial containment pressure, Core Makeup Tank (CMT) resistance multiplier and initial containment humidity.
Development of TMSR-SF1 Protection System Prototype
Authors: Hou Jie , Liu Ye , Li Bingying , Wei Yongbo , Huang Guoqing , Liu Guimin
Published: October 2017

The reactor protection system (RPS) of TMSR-SF1, which is the first 10MW Solid Fuel experiment reactor for Thorium based Molten Salt Reactor (TMSR) Energy system, adopt FPGA as the key technology to develop the system based on the triple modular redundancy system architecture. The RPS prototype was developed follow the TMSR-SF1 conceptual design to demonstrate the function and performance and to test the system reliability. The system design of TMSR-SF1 RPS is introduced in detail in this paper, the emergency trip parameter is also discussed. The hardware & software design of the prototype including the kernel logical board card design, the chassis and cabinet design, the breaker cabinet design, etc. are described. The prototype passed the performance test and the third-party testing, results and discussion are reported.
Seismic Margin Assessment and Improvements of Qinshan Nuclear Power Plant
Authors: Xu Leilei , Zhao Xiaoling , Li Zhaohua
Published: October 2017

Seismic Margin Assessment (SMA) is one of the methods for seismic safety assessment of nuclear power plants. The United States in the 1990s requires all running nuclear power plants to carry out an Individual Plant Examination of External Events (IPEEE), in the completion of the IPEEE of 110 units, 65 units using the seismic margin assessment method. The commercial nuclear power industry of China started late, although the Shanghai Nuclear Engineering Research and Design Institute in the 1980s to carry out seismic safety evaluation of relevant research. After the Fukushima nuclear accident, the National Nuclear Safety Administration requires all nuclear power plants in operation to carry out Seismic Margin Assessment. In view of the above background, the Shanghai Nuclear Engineering Research and Design Institute and Qinshan Nuclear Power Co., Ltd. on the Qinshan nuclear power plant to carry out the seismic margin assessment. Through the assessment, some weak links were found in the Qinshan nuclear power plant in some aspects. The Qinshan nuclear power plant in the implementation of the targeted improvements, the power plant’s seismic capacity has been effectively improved.
The Study of CAP1400 Core Melt and Relocation for IVR Analysis
Authors: Wang Jiayun , Lu Wei , Gu Pei Wen
Published: October 2017

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.
Experiments on the Splashing Limit During Drop Impact Onto a Thin Liquid Film
Authors: Kitabayashi Sota , Enoki Koji , Okawa Tomio
Published: October 2017

The phenomenon of secondary droplet production during single drop impingement onto a liquid film is encountered in many industrial situations. Typical examples in the field of nuclear engineering are the spray cooling of hot surface and the atomization of radioactive liquids in severe accident. Therefore, the prediction of the onset of secondary droplet production is very important. It is known that the two types of droplet splashing mechanisms are present: the prompt splash and the late splash. The main purpose of this research is to determine the splashing limit separately for the prompt splash and the late splash. It is expected that the splashing limits are expressed using the three dimensionless numbers: the Weber number, the Ohnesorge number, and the dimensionless film thickness. Experiments were hence carried out using pure water and silicone oil as the working liquid. The experimental ranges were 129–606 for the Weber number, 0.00183–0.00300 for the Ohnesorge number, and 0.13–3.0 for the dimensionless film thickness. It was found that the occurrence of splashing can be predicted more accurately if the splashing limit is evaluated separately for the prompt splash and the late splash.
Air Velocity Influence on Heat Transfer in the PCCS
Authors: Yang Lin , Li Cheng , Du Wang-Fang , Gao Zhan , Zhou Shan
Published: October 2017

The passive containment cooling system (PCCS) is important passive safety systems in the Advanced Pressurized Water Reactor (APWR), which belongs to the generation III of nuclear reactors. In design basis accident (DBA), the steam condenses on the inner surface of the containment shell and the cooling water evaporates from the outer surface of the containment shell. In this process, the heat is transferred from the inside of the containment to the outside. To span the expected range of conditions and provide a proper model for evaluating the inner steam condensation coupled with outer evaporation heat transfer process, the inner steam condensation coupled outer evaporation experimental test (ISCOE) is developed by State Nuclear Power Technology Research & Development Centre (SNPTRD). Several tests have been done on the ISCOE experimental test facility. The influence of different key factors for the capability of the heat transfer of the containment steel shell wall has been researched. Key factors include steam pressure, steam temperature, water film velocity, air velocity, steel shell wall angle, and so on. The result of these tests has an important significance to the research of heat transfer capability of the containment steel shell wall. In this paper, several tests are introduced, including details, results and analysis. The influence of air velocity for the capability of the heat transfer of the containment steel shell wall is also analyzed.
Simulation of Radiation Monitoring Instruments Data for Emergency Response Exercises
Authors: Wang Liu , Feiye Liao , Pingting Jiang , Dongyu He , Yong Luo , Yehong Liao
Published: October 2017

Radiation monitoring instruments (KRT) is important to decide emergency response level (EAL) in accident situation. Emergency response drills is more and more significant after Fukushima Daiichi severe accident. This article develops a code to simulate radiation monitoring instruments data for emergency response exercises. A part of calculated input of the code comes from MAAP calculated results included source terms and thermo hydraulic data. KRT simulation code runs with MAAP calculation results and output KRT simulated value at the same time. This article gives a way to prove the result of KRT simulation code analyzed is correct and matched through simulating an emergency response exercise scenario.
Effects of Annealing on the Corrosion Property of CZ1 Alloy
Authors: Chen Liutao , Tan Jun , Gao Changyuan , Wen Dungu , Zou Hong , Xu Yang , Wang Xu
Published: October 2017

CZ alloys were developed by China General Nuclear Power Group (CGN) for pressurized water reactor. To improve the corrosion resistance, the impacts of annealing on corrosion of CZ1 alloy were studied. The fuel cladding tubes of CZ1 alloy was fabricated by different annealing processes, with intermediate annealing in the range of 560 °C to 700 °C. Then the corrosion properties of CZ1 alloy and low Tin Zr-4 specimens were investigated by autoclave tests. The matrix and second phase particles along with the oxide characteristics were analyzed. The experimental results showed that the corrosion resistance of CZ1 alloy is superior to that of low Tin Zr-4,. Besides, Lower corrosion weight gain is obtained in the cladding tubes with decreasing annealing temperature, which correlates well with the smaller size of second phase particles. The underlying mechanism was discussed. The results in this study will be used as a guide for the material processing design of CZ1 alloy.
Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor
Authors: Ge Jian , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: October 2017

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.
Numerical Simulation of Graphite Dust Deposition in Pebble Bed Reactor Core of HTGR
Authors: Chen Tao , Wang Jie , Peng Wei , Sun Xiaokai
Published: October 2017

The deposition of graphite dust produced by the collision between graphite components would cause the security issue in HTGR and need to be analyzed. In this paper, a numerical calculation about the graphite dust’s deposition in the pebble bed reactor core of high-temperature gas cooled reactor was conducted. The three-dimensional steady-state solver was employed for the calculation of flow field and temperature field during simulation. The discrete phase model (DPM) and Lagrange method were applied for the simulation of graphite dust. Effects of parameters such as particle diameter, pebble bed layer numbers, inlet velocity and surface temperature of fuel elements on deposition of graphite dust are analyzed. The results indicated that a majority of particles deposit on the first layer pebble because of first layer’s shielding effect on nether layers. Moreover, deposition efficiency of graphite dust increases with increasing particles diameter and increasing inlet velocity due to greater motion inertia of particles. Compared with fewer layers, more layers structure would lead to larger deposition efficiency because of more opportunities for collision between graphite dust and pebbles, but the difference is not obvious. In addition, the higher surface temperature of fuel elements would cause lower deposition efficiency due to larger thermophoretic force which would drive particles to deviate from pebbles.
Penetration Jet Impingement Calculation in Engineering Application
Authors: Wu Tong , Zhou Ying , Qi Tao
Published: October 2017

In the process of nuclear power plant design, Pipe Rupture Hazards Analysis (PRHA) was obliged, including postulated rupture location and configurations, jet impingement effects, compartment pressurization effects, environmental influences, flooding effects, leak-before-break and influence on SSC, etc. The analysis of jet impingement is of great importance; aims at obtaining the jet impingement configuration and the impingement force acting on the target. Jet impingement configuration and force depended on the jet flow properties. A jet discharging from a saturated steam line will accelerate and expand due to the pressure differential, and it will partially condense to a low-moisture wet steam with liquid phase in the form of dispersed, entrained water droplets. A jet discharging from a sub-cooled or saturated hot water line (greater than 100°C) would flash to a low quality wet steam and the flashing would cause the jet diameter to expand very rapidly. These jet flows have a phase change and two-phase flow process; the recommended two-phase flow model that should be used was presented in ANSI/ANS 58.2-1988. However, penetration jet impingement which is often encountered in the PRHA was not introduced. In normal cases, the saturated steam, sub-cooled or saturated hot water (greater than 100°C) expands directly to the surroundings. But for penetration jet impingement, the fluid is first discharged to the narrow annular section formed by the pipe and the penetration, then flows through this area, and finally expands to the surroundings at the open side of the penetration. The penetration expanding jet analysis is much more complicated. A method of determining penetration jet impingement in engineering applications was derived based on the fundamental method presented in ANSI/ANS 58.2-1988, Henry and Fauske model recommended. The simplified method took advantage of the two-phase flow models and equations given by ANSI/ANS 58.2-1988; the jet configuration could be calculated effectively and the target impingement force could be derived using the result presented by these equations simultaneously. The impingement pressure field was defined using the program for different initial states of the postulated pipe rupture — namely sub-cooled and saturated. The pressure distribution along the jet centerline obtained has shown clearly the three regions in ANSI/ANS 58.2-1988. The pressure field has shown that sub-cooled water has a larger zone of influence and saturated vapor has a higher mean impingement pressure as Sub-cooled water was under expanded while saturated vapor has higher enthalpy as it contains more energy.
Study of Flow and Heat Transfer of Low Power Natural Circulation Under Rolling Condition
Authors: Ji Peng , Gao Puzhen , Zhang Yinxing
Published: October 2017

Experimental and numerical studies on the flow and heat transfer characteristics of low heat flux natural circulation under the rolling condition were conducted in this paper.

The flow loop applied in the experiment is fixed on a rolling platform and the test fluid is deionized water. The testing channel made of stainless steel is narrow rectangular with the cross section of 2×40 mm2 and the heat length is 1100 mm. Enough experimental data points are obtained for a system pressure range of 0.1–0.3 MPa and an inlet temperature range of 50°C∼80°C. The testing channel is heated by electric, and the heat flux is less than 60 kW/m2. The rolling platform is driven by the crank rocker mechanism, and the rolling angle of the experimental loop varies with sine function. Driven by low pressure head, the low heat flux natural circulation has a negative flow value in the trough because of the effect of additional pressure drop under rolling condition. The experimental results show that the temperature fluctuation of the coolant is more severe because of intermittent reverse of the flow, and the parameter of outlet water will turn high under the effect of repeated heating. At the same time the amplitude of the temperature fluctuation in the outlet and inlet increase sharply compared with the condition that the flow doesn’t reverse. Under this condition the temperature of water increases as the additional inertia force raises because the level of flow reverse is improved and the working fluid is heated for more than once.

The heat transfer of low power natural circulation under rolling condition is discussed in this paper. The analysis finds that the rolling condition doesn’t strengthen the heat transfer absolutely when the heat power is low. The average heat transfer coefficients of the natural circulation under rolling conditions increase faster than those under steady state when strengthening the power of heat. Considering the flow will reverse periodically, the reason why the heat transfer is weakened at low power condition is given.
Cyber Security Assessment of Component Off-the-Shelf Based NPP I&C System Using IMECA Technique
Authors: Zelinko Ilona , Kharchenko Vyacheslav , Leontiev Konstantin
Published: October 2017

Nowadays cyber security assurance is one of the key challenges of safety critical software based NPP I&C (Nuclear Power Plants Instrumentation and Control) systems requirements profiling, development and operation. Any I&C system consists of a set of standard software (SW), hardware (HW) and FPGA components. These components can be selected and combined in different ways to address the particular control and safety assurance related tasks. Some of them are proprietary software (PS) and commercial off-the-shelf (COTS) components developed previously. Application of such components reduces the level of safety and cyber security, because they can contain vulnerabilities that were created intentionally. In this case, targeted attacks can lead to a system failure.

National Vulnerability Database (NVD) and other open databases contain information about vulnerabilities which can be attacked by insiders or other intruders and decrease cyber security of NPP I&C systems.

In this paper, we propose a safety assessment technique of NPP I&C systems, which consists of the following procedures:

1. Analysis of I&C architecture to assess influence of OTS component failures on dependability (reliability and safety) of the system. For that purpose, FMEDA or similar techniques can be applied.

As a result, three-dimension criticality matrixes (CM) (with metrics of detection, probability and severity) are developed for different components (SWFCM and HW/FPGAFCM).

2. The IMECA-based assessment of OTS components and their configuration. In this case, CMs (SWICM and HW/FPGAICM) describe the degree of failure component influence on cyber security.

3. Joining of criticality matrixes (SWFCM and HW/FPGAFCM, SWICM and HW/FPGAICM), impact analysis of components depending on degree of influence on cyber security and safety as a whole.

4. Developing of Security Assurance Case and selecting of countermeasures according to safety (cyber security)/costs criteria.

The developed tool supports creation of criticality matrixes for each analyzed component of the system and I&C as a whole. Joining of criticality matrixes allows creating common matrix for system cyber security and functional safety. The tool supports decision making to optimize choice of countermeasures according to criterion of safety and security/cost criterion.
Effects of Heater Material and Surface Orientation on Heat Transfer Coefficient and Critical Heat Flux of Nucleate Boiling
Authors: Mei Yong , Zhu Yechen , Zhang Botao , Gong Shengjie , Gu Hanyang
Published: October 2017

External reactor vessel cooling (ERVC) is the key technology for In-Vessel Retention (IVR) to ensure the safety of a nuclear power plant (NPP) under severe accident conditions. The thermal margin of nucleate boiling heat transfer on the reactor pressure vessel (RPV) lower head is important for ERVC and of wide concern to researchers. In such boiling heat transfer processes, the reactor vessel wall inclination effect on the heat transfer coefficient (HTC) and critical heat flux (CHF) should be considered. In this study, experiments were performed to investigate the effects of heater material and surface orientation on the HTC and CHF of nucleate boiling. Copper and stainless steel (SS) surfaces were used to perform boiling tests under atmosphere pressure. The orientation angle of both boiling surfaces were varied between 0° (upward) and 180° (downward). The experimental results show that the surface orientation effects on the HTC is slight for both the copper surface and the SS surface. In addition, the relationship of measured CHF values with the inclination angles was obtained and it shows that the CHF value changes little as the inclination angle increases from 0° to 120° but it decreases rapidly as the orientation angle increases towards 180° for both boiling surfaces. The material effect on CHF is also observed and the copper surface has higher CHF value than the SS surface. Based on the experimental data, a correlation for CHF prediction is developed which includes both the surface orientation effect and the heater material effect.
Numerical Reproduction of DDT in Small Scale Channels
Authors: Ren Ke , Kotchourko Alexei , Lelyakin Alexander , Jordan Thomas
Published: October 2017

Deflagration to detonation transition (DDT) is a quite challenging subject in computational fluid dynamics both from a standpoint of the phenomenon nature understanding and from extremely demanding computational efforts. In the article the hybrid DDT combustion model is introduced as an efficient method to simulate the DDT problems. As verification, two DDT experiments made in experimental facility MINI RUT are used.
Popularization and Application of a New Flexible Shielding Material in Nuclear Power Plant
Authors: Yulong Li , Feng Liu , Zheng Kang , Jie Zhan , Xiajie Liu
Published: October 2017

Excessive gamma rays will be emitted when a nuclear power plant is under the refueling overhaul, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding material with a “sandwich structure” was developed. This kind of material has been adopted in a mega-kilowatt pressurized water reactor (PWR) nuclear power plant as the first demonstration project. Based on the characteristics of the new flexible shielding material, the process design program of the product application was founded, and the new material was compared with the traditional lead shielding material. At last, the installation and removal time together with the shielding effect were analyzed according to the first demonstration application.
Application of GIS in the General Layout of a Nuclear Power Plant
Authors: Fu Peng , Wang Ming
Published: October 2017

The general layout and design of a nuclear power plant is an important component of its design. This study purports to explore the feasibility of developing related geographical database management technology and applying specialized analysis tools for GIS by using 3-D modeling technology and taking engineering requirements of the layout and design of a nuclear power plant into consideration. A case study has been developed which introduces GIS database management and application based on an actual domestic nuclear power plant. By comparison and optimization of the GIS application system, the study proved that this method can assist planners in designing nuclear power plants, and ultimately improves the safety and efficiency of nuclear power plant construction and operation.
In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches
Authors: Shen Pei , Zhou Wenzhong
Published: October 2017

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.