bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Probabilistic Risk Assessment Methodology of Decay Heat Removal Function Against Combination Hazards of Strong Wind and Rainfall for Sodium-Cooled Fast Reactors
Authors: Yamano Hidemasa , Nishino Hiroyuki , Kurisaka Kenichi
Published: October 2017

This paper describes probabilistic risk assessment (PRA) methodology development against combination hazard of strong wind and rainfall. In this combination hazard PRA, a hazard curve has been evaluated in terms of maximum instantaneous wind speed, hourly rainfall, and rainfall duration. A scenario analysis provided event sequences resulted from the combination hazard of strong wind and rainfall. The event sequence was characterized by the function loss of auxiliary cooling system, of which heat transfer tubes could crack due to cycle fatigue by cyclic contact of rain droplets. This situation could occur if rain droplets ingress into air cooler occurs after the air cooler roof failure due to strong-wind-generated missile impact. This event sequence was incorporated into an event tree which addressed component failure by the combination hazard. Finally, a core damage frequency has been estimated the order of 10−7/year in total by multiplying discrete hazard frequencies by conditional decay heat removal failure probabilities. A dominant sequence is the failure of the auxiliary cooling system by the missile impact after the failure of external fuel tank by the missile impact. A dominant hazard is the maximum instantaneous wind speed of 40–60 m/s, the hourly rainfall of 20–40 mm/h, and the rainfall duration of 0–10 h.
A Movable Device of Explosives Detection
Authors: Yan Zheng , Ti Tong , Shi-wei Jing , Huan-yu Li , Xuan Liu , Ya-dong Gao , Rui-xue Nian
Published: October 2017

This paper introduces a movable detection device for explosives hidden in wall. The device is composed of the 14MeV neutron generator, the silicon detector and Φ76mm×76mm LYSO detector. The γ spectra produced by the inelastic scattering interaction provide the relative counts of C, N, O, which are converted into chemical fractions to distinguish the explosives from the organic compounds. The device using associate particle method reduces the background 200–300 times. The 300g ammonium nitrate and TNT explosive simulated samples placed behind the 15cm-thick and 10cm-thick wall is detected by the device, respectively, when the neutron intensity is 1×107n/s. Compared with other methods, the experimental results show that the device with associate particle method has higher detection space resolution ability and explosives recognition ability. The device can provide the support for the detection of explosives.
The Development and Application of the Engineering and Design Analyzer of CAP1400 Nuclear Power Plant
Authors: Yuanlei He , Qijiang Zhang , Xiaoyan Li
Published: October 2017

In the process of research and development for a new nuclear power plant, it is very necessary to develop a dynamic platform and tools to analyze and verify the plant control & protect system and human factor engineering. Therefor, Shanghai Nuclear Engineering Research and Development Institute (SNERDI) developed the Engineering & Design Analyzer of CAP1400 Nuclear Power Plant (CAP1400 EDA) which provides a dynamic platform environment for analyzing and verifying the control system and human factor engineering of the CAP1400 nuclear power plant, a new Gen III passive nuclear power plant.

In this paper, the mechanism and implement approach of the CAP1400 EDA will be mainly introduced, for example, the platform architecture of the EDA, analysis tools integrated in the EDA and CAP1400 nuclear power plant modes based on the EDA.

In the meantime, a typical application case based on the CAP1400 EDA will be demonstrated in this paper, for example the capability of the NSSS control system will be verified in a ramp load down & raise operate transient. In this transient process, the NSSS control system of the plant is assessed whether it has the capability to keep the key parameter and state of the plant in an acceptance condition or range. And also other transients such as step load transient, large load transient can be simulated on CAP1400 EDA to verify whether or not the NSSS control systems are properly designed.
Research on Severe Accident Management Strategies of Spent Fuel Pool
Authors: Pingting Jiang , Chunrong Liu , Wang Liu , Dongyu He , Juanhua Zhang
Published: October 2017

It is more and more concerned after Fukushima Daiichi about spent fuel pool severe accident management, and it is also required by the regulatory body of NSC (Nuclear and Radiation Safety Center). During such a beyond design basis event, spent fuel assembly becomes over-heated and finally damaged. In order to prevent and stop SFP (spent fuel pool) severe accident process, strategies should be developed when existing procedures are no longer effective or available. This paper gives a development approach of CPR1000 NPP (nuclear power plant) SFP severe accident management guidelines. According to this development approach, the framework of SFP severe accident management guidelines has been developed. To fulfill the strategies and framework for SPF severe accident guidelines, calculation of typical accident cases and determination analysis are carried out using PSA method, based on investigation and survey of international severe accident management procedures research status, considering the characteristics of CPR1000 NPP and SFP design. Then related actions and proper times are proposed, which would help to develop detailed severe accident management guidelines (SAMG) of SFP.
Adaptive Mesh Refinement for Neutron Transfer With Lattice Boltzmann Scheme
Authors: Ma Yu , Wang Yahui , Song Kuilong , Sun Qian
Published: October 2017

This paper presents a novel lattice Boltzmann (LB) model for neutron transfer and a block-structured adaptive-mesh-refinement (SAMR) technique for proposed LB model. By discretizing the general Boltzmann equation, the LB model for neutron transfer is established and the corresponding parameters are obtained. The SAMR technique removes the requirement of tree-type data structure in traditional adaptive-mesh-refinement technique and adjusts the time step adaptively and identically in all blocks. By applying the node-type distribution function, the needs for rescaling the distribution functions is eliminated. To solve the discontinuities of scalar flux at fine-coarse blocks interface, a novel technique is presented which treats the inner boundary condition by streaming process of LB method. Simulation results show good accuracy and efficiency of the proposed neutron LB model with SAMR technique. This paper may provide a powerful technique for large engineering calculation.
Data Assimilation for Fukushima Nuclear Accident Assessments
Authors: Xu Mei , Yuan Biao , Wang Liangyu , Zhang Lijun
Published: October 2017

In order to investigate the feasibility of data assimilation in a real nuclear accident environment, measurements of Fukushima nuclear accident were considered. The data assimilation system was constructed by using the Lagrangian puff model as the radioactive material diffusion model, and 86 group real dose rate data from the accident as the observations, and the Ensemble Kalman Filter algorithm as the assimilation algorithm. The experimental results show that the assimilated nuclear accident radiation field is in good agreement with the actual measurements, the land contaminated areas are concentrated in the northwest of the nuclear power plant. With the increase of the real measurements, the error of the radiation field decreases with time. Compared with the results with no assimilation, the uncertainty of assimilated dose rate was reduced more than 80%. Through the data assimilation, the whole error of the radiation field is about 30%. The utilization of the real measurements can reduce the uncertainty of the model prediction.
Experimental Study of the Sodium-Cooled Fast Reactor Vessel Seismic Analysis Considered the Liquid Sloshing
Authors: Liu Hongda , Lu Daogang , Li Xiaoxuan , Zeng Xiaojia , Bao Guangdong
Published: October 2017

Sodium-cooled fast reactor vessel is one of seismic grade I equipment, so the vessel seismic analysis considered the liquid sloshing is an important topic in fast reactor vessel design. In order to investigate the behavior of liquid sloshing under the condition of seismic, a fast reactor vessel model at 1:25.7 scale was designed. Using water simulate liquid sodium, we carried out modal experiment and seismic table test. By using hammering test measured natural frequency of model structure under the condition of water and without water; By utilizing seismic table frequency sweep test measured frequency of water sloshing in model; Then, input water sloshing sine wave of first order frequency and second order frequency respectively through seismic table.After that measure wave height and hydraulic pressure response of liquid sloshing. The experimental results show wave height and hydraulic pressure response of liquid sloshing in second order frequency is bigger than the first order frequency. This conclusion can provide favorable reference for the seismic design of the sodium-cooled fast reactor.
Study on TlBr Radioactive Detector
Authors: Li-xia He , Xiao-yong Hao , Gao-kui He
Published: October 2017

TlBr is a kind of semiconductor material. Due to its promising physical properties and can be used at room temperature, it is continually studied as X and gamma ray detectors candidate material. Both of its atomic number and density are high. It also has large band-gap (B = 2.68eV) and low ionization energy. TlBr device exhibits high detection efficiency and excellent energy resolution. It can be easily fabricated or compacted in small housing. So it is a reasonable selection in the fields of nuclear material inspection and safeguards property, national security, spatial and high energy physics researches. The paper investigates the TlBr radioactive detector development and fabrication procedures. The processing detail information and signals collection are emphasized in different section. The prototype detectors were irradiated by Am-241 and corresponding spectrum was obtained. The photoelectric peak at 59.5keV is distinguished visible and the best resolution at 59.5keV is 4.15keV (7%).
Multi-Dimensional Shield Performance Analysis Through an Even Parity Based “Contributon” Transport Concept
Authors: Yousefi Mostafa , Zolfaghari A. , Minuchehr A. , Abbassi M. R.
Published: October 2017

In this paper we present an even parity approach for deriving the detector response flux. Forward and adjoint angular fluxes, ψ(r,Ω) and ψ†(r,Ω) obtained from the K+[ψ+] and K†+[ψ†+] variational principles are used to determine the spatial channels in shielding materials. These spatial channels clearly show the dose rate passages from the source to the detecting point, hence weak and strong points of the shielding design is illustrated thereafter. Spatial and angular components of the fluxes are approximated using the finite element method (FEM) as well as the spherical harmonics polynomials (SHP), respectively. By the adjoint weighted even parity flux, we can obtain the multigroup response fluxes over arbitrary shaped multidimensional geometries with less computational efforts compared to full parity approaches. A number of test are examined via the code ENTRANS, developed for Even parity Neutron TRANSport calculations. Results confirm ability and robustness of the proposed approach.
Improvement of the Reactor Control System in AP1000
Authors: Chunhui Dai , Xinyu Wei , Mo Tao , Jun Wu , Zhenxing Zhao , Zhouyang Liu , Yong Liu
Published: October 2017

AP1000 is a two-loop pressurized water reactor (PWR) developed by Westinghouse, which uses the forces of nature and simplicity of design to enhance plant safety and reduce construction costs. Compared to traditional reactor control strategies, the AP1000 adopts an advanced Mechanical Shim (MSHIM) core control strategy, which controls the reactivity and the axial power distribution without the need of boron adjusting. The main control signal of MSHIM is the difference of average core coolant temperature and reference temperature obtained from turbine load. In order to improve the response speed, a mismatch signal obtained by the difference of the nuclear power and target power is introduced as the auxiliary control signal. However, this auxiliary signal may cause the control rod acting too fast, also, the power mismatch channel noise is not suitable as a differential lead control channel. To solve this problem, we remove the nuclear power signal channel, and use the average coolant temperature error as the only control signal. Furthermore, a P controller is added for adjusting the control signal to increase the response speed and reduce the steady error. The proposed method was verified by dynamic performance simulation and compared to the original control system. The calculation results show that the modified power control system could not only simplifies the control logic but also enhances the control performance. And, the proposed method satisfies the needs of reactor core power control and power distribution control. The conclusions have reference value for the engineering practice.
Study on the Indication Processing Method for the In-Service Inspection Based on RSE-M
Authors: Peng Xia , Xueliang Zhang , Cheng Yang , Qingyue Zeng
Published: October 2017

For the defect examined by non-destructive testing during in-service inspection, it shall be subject to a set of operations known as indication processing, which aim to identify and analyze a deviation situation, confirm whether or not this is a genuine defect and select the best processing solution. This paper is intended to illustrate an indication processing methodology that provides guidance on performing indication processing. Firstly, an indication processing method which covers indication identifying, deviation analysis, solution assessment, and solution implementation is specified. And then, some key technologies such as defect characterization, fatigue crack growth analysis applied in the indication processing methodology are descripted. Finally, some preliminary thinking is presented during introducing the indication processing methodology and some technical suggestions are also provided to perfect the methodology for providing better reference during application on PWRs in China. Actually, the study in this paper is preliminary and more research is needed to be done in the future.
Design and Optimization of Temperature Acquisition System for Determination of Effective Thermal Conductivity of Pebble Bed
Authors: Cheng Pengxin , Ren Cheng , Wu Yongyong , Li Rui
Published: October 2017

A full-scale heat transfer test facility has been designed and built for the determination of effective thermal conductivity of pebble bed, which is a macroscopic parameter to characterize the heat transfer capacity of the core in the High Temperature Gas-Cooled Reactor. The data acquisition system is developed to collect, display and record the temperature data in monitoring points. Two alternative software systems are designed to obtain better performance. To enhance precision of the measurement system, several aspects are analyzed and optimized in the implementation of LabVIEW. The error of the hardware system is analyzed, which is within the acceptable range. The data acquisition system can meet the practical demands of temperature acquisition in the range of thermal analysis.
Similar but Different: A Top-Down Benchmarking Approach to Investigate Nuclear Decommissioning Projects
Authors: Invernizzi Diletta Colette , Locatelli Giorgio , Brookes Naomi J. , Grey Martin
Published: October 2017

Project management literature has, until now, mainly focused on new build and only in the last decades the issues of decommissioning (mega) projects has arisen. To respond to this changing environment, project management will need to understand the challenges of decommissioning projects. Decommissioning projects within Oil & Gas, Chemical and Nuclear sectors are characterized by high costs, long schedules and uncertainty-based risks. The budget for Nuclear Decommissioning Projects and Programmes (NDPs) are subject to well publicized increases and, due to their relatively recent emergence, complexity and variety, key stakeholders lack a full understanding of the key factors influencing these increases.

Benchmarking involves “comparing actual or planned practices [...] to identify best practices, generate ideas for improvement” [1] and offers significant potential to improve the performance of project selection, planning and delivery. However, even if benchmarking is the envisaged methodology to investigate the NDPs characteristics that impact on the NDPs performance, until now, it has only been partially used and there is a huge gap in the literature concerning benchmarking NDPs.

This paper adapts a top-down benchmarking approach to highlight the NDPs characteristics that mostly impact on the NDPs performance. This is exemplified by a systematic quantitative and qualitative cross-comparison of two major “similar-but-different” NDPs: Rocky Flats (US) and Sellafield (UK). Main results concern the understanding of the alternatives of the owner and/or the contractors in relation to (1) the physical characteristics and the end state of the nuclear site, (2) the governance, funding & contracting schemes, and (3) the stakeholders’ engagement.
Design and Implementation of Visual Design Datacenter for Nuclear Power Plant
Authors: Hui Lu , Zhong Liu , Bo Niu , Lin Zhang , Hui Rong , Chunping Hu , Yu Liu
Published: October 2017

With the rapid development of information technology, and the projects are increasingly large, complex, and then the flaws gradually appear only by using CAD design software in the projects. Although the new engineering software has many new features, it can greatly improve design quality and efficiency, the lack of data exchange between the software, so it’s easy to form Data Silos. To solve inconsistencies and other issues caused by Data Silos, the article designs and builds a design data integration management and analysis system by JavaEE technologies and SPF.
The Comparative Advantages of CAP1400 Nuclear Power Plant Passive Safety System
Authors: Cheng Ye , Minglu Wang , Zhongming Qiu , Yong Wang
Published: October 2017

With the demand for nuclear power increasing, the first choice of almost all countries who want to build a new nuclear power plant is to use generation III technology, primarily because the safety of generation III technology is greatly improved compared with that of generation II and II + technology. The passive safety technology was introduced by the AP1000 and is one of the best applications of generation III technologies. In this study, the representative passive containment cooling system of the CAP1400 (developed based on AP1000) and the containment spray system of a generation II nuclear power plant are compared and analyzed using the Probabilistic Safety Assessment method. The reasons why a passive safety system has comparative advantages are determined by concrete calculations.
Analysis of 3-Dimensional Hydrogen Behavior With Passive Containment Cooling
Authors: Huang Xingguan , Fu Tingzao , Huang Gaofeng
Published: October 2017

Hydrogen safety is one of the most important issues for severe accident analysis. Comparing with the lumped parameter code, 3-D CFD code GASFLOW has more advantages for analyzing hydrogen related phenomena [1]. However, due to the lacking of passive containment cooling system (PCS) model, GASFLOW is inapplicable for hydrogen analysis of PWR with PCS. This paper shows the development procedure of PCS model, including the model introduction and validation. The main functions of the PCS model are the calculation of falling water film thickness, heat transfer between wall and film, and film evaporation. Then, the hydrogen safety analysis by GASFLOW with PCS model is performed for CAP1400.
Research on Detailed Depletion Chain Solving Algorithm With the NUIT Code
Authors: Li Jian , She Ding , Shi Lei
Published: October 2017

Burnup calculation and nuclide inventory calculation may require solving detailed depletion chains with high computational accuracy and efficiency. NUIT (NUclide Inventory Tool) is a newly-developed nuclide inventory calculation code, which is capable of handling detailed depletion chains by implementing various advanced algorithms. Based on the NUIT code, this paper investigates the accuracy and efficiency of the Mini-Max Polynomial Approximation (MMPA) method, and compare it with the Chebyshev Rational Approximation Method (CRAM). It is concluded that both methods are numerically accurate and efficient for dealing with detailed depletion chains with extremely short half-lived nuclides.
Analysis on Impact of Fuel Thermal Conductivity Degradation (TCD) on Large Break Loss of Coolant Accidents of CAP1000
Authors: Weiwei Wang , Lu Lu
Published: October 2017

Under high burnup conditions, thermal conductivity of fuel pellet degrades, which is referred to as thermal conductivity degradation (TCD). TCD phenomenon influences fuel average temperature and fuel storage energy under steady state condition before loss of coolant accident (LOCA) and further influences peak cladding temperature (PCT) during large break LOCA process. In this study, sensitivity study on double ended guillotine break of cold leg in CAP1000 at different burnup conditions was performed, using large break LOCA analysis code WCOBRA/TRAC and PCTs under different conditions were obtained. The modified NFI (Nuclear Fuels Institute) TCD model was adopted to model fuel conductivity after degradation in analysis and decrease of peaking factors including FQ and FΔh after 30GWD/MTU was also considered. Sensitivity analysis showed that: after considering the influence of TCD and peaking factor burndown, the PCT limiting case did not occur in low burnup range again, but occurred at burup of about 29GWD/MTU. Compared to other burnup points, the first and second peak values of PCT at that burnup point were all at the highest level. Performing of this study could prefer reference for analysis and estimation of large break LOCA of passive nuclear power plants under high burnup conditions.
Knowledge From Recent Investigation on Local Fuel-Coolant Interaction in a Liquid Fuel Pool
Authors: Zhu Ting , Cheng Songbai , Lin Shaopeng
Published: October 2017

Studies on local Fuel-Coolant Interactions (FCI) in a liquid pool are of crucial importance for the improved evaluation of severe accidents for Sodium-cooled Fast Reactors (SFR). To clarify the mechanisms underlying this interaction, several years ago a series of experiments was performed by Cheng et al. (2014) at the Japan Atomic Energy Agency (JAEA) through delivering a given quantity of water into a simulated molten fuel pool which is formed by a low-melting-point alloy. In order to acquire more evidence, numerical analysis using the SIMMER-III, an advanced fast reactor safety analysis code, was also conducted. However, through those analyses, many limitations of the experiments (esp. the uncertain initial geometry of water lump) which would impair the reliability of experimental observations and prevent the direct comparisons with calculations were realized. Focusing on those aspects, further investigations from both experimental and numerical aspects were continued at the Sun Yat-sen University in China. In this study, the effect of injection mode (namely the coolant-injection mode and fuel-injection mode) on local FCIs is studied using the FLUENT code along with its Volume of Fluid Model (VOF). It is seen that the interaction mode does have some influence on the transient behaviors, in particular the formation of steam bubbles, and time variation of temperature and pressure values. In addition, the difference on heat exchange rate due to the steam bubble formation is also confirmable. Knowledge and evidence from this study might be utilized for future development and analyses of SFR severe accident analysis codes in China.
Splash During Liquid Jet Impingement Onto a Horizontal Plate
Authors: Zhan Yi , Oya Naoki , Enoki Koji , Okawa Tomio , Ohno Shuji , Aoyagi Mitsuhiro , Takata Takashi
Published: October 2017

In the numerical analysis of sodium fire event following coolant leakage in a sodium-cooled fast reactor, sufficient understanding is needed for the liquid jet atomization to estimate the combustion rate accurately. In the present work, a liquid jet was emanated vertically downward from a circular nozzle onto a liquid film formed on a horizontal plate. The droplets produced at the impact point were investigated. The splash ratio (the ratio of the mass of the splashed droplets to the jet flow rate) was measured under varied experimental conditions of nozzle diameter, fall height and flow rate of liquid jet. The experimental result depended significantly on the morphology of liquid jet upon impact. The splashing rate was negligibly small when the liquid jet impinged as the continuous jet but a significant amount of liquid was splashed when the liquid jet impinged as the broken jet. Thus, we developed a method to estimate the impact frequency. It was shown that the splash ratio can be correlated well if the impact frequency is included in the correlation.
Risk Analysis of Water Intake Structure Blockage in a Nuclear Power Plant
Authors: Wang Jinkai , Duan Feng
Published: October 2017

With the increasing human activities and natural environment changes in the vicinity of nuclear power plants, intake structure blockage has become a major risk to challenge the nuclear power plant operation safety. This risk has been ignored for long time. There are a lot of external events that may cause intake structure blockage, such as biological events, floating objects, sand sedimentation, oil spilling, etc. The root causes to form these events are very complex and they are also very site-specific. As these external events occurrence frequencies are very difficult to be predicted due to very limited acknowledgement, current practice seldom uses detailed quantitative method to evaluate the intake structure blockage risk. This study proposed a probabilistic method in this risk analysis. The process includes external events identification and screening, plant response and quantification. A nuclear power plant was selected in this analysis as a case study. This method took a systematic process and identified some external events that were dominant contributors to the intake structure blockage. The result indicated that intake structure blockage caused by external events could be dominant risk contributors to the plant and their risk should not be ignored. Some recommendations were provided to reduce the risk. Engineering practice proves that this method is very useful to identify the plant weak points for the external events and can help to improve the plant safety.
Characteristics of Thermal-Hydraulic and Heat Transfer in Liquid Windowless Target of ADS
Authors: Wang Feng , Wen Qiang , Qin Xue
Published: October 2017

In recent years, the morphological characteristics and stabilization methods of free interface in liquid windowless target become hot research topics in accelerator driven subcritical system (ADS). Based on the structure design of a certain windowless spallation target, CFX program was used to simulate and analyze its free interface character. The method of k-ε turbulence model, Cavitation model and VOF method were used to study the flow characteristic of liquid lead-bismuth alloy (LBE) with cavitation phase change and to analyze the free interface morphology characteristics of coolant in the target area. It is concluded that the target region form two stable free interfaces when fluid outlet pressure is in the range of 10000–40000Pa and fluid entrance velocity is in the range of 0.5–1.2m/s. The flow field near the free interface structure is complex. The vortex region appears, and the disorders in the vortex flow pattern leads to fluctuation of the free interface. After establishing stable free interface morphology, the heat transfer characteristics of windowless target was further analyzed.
Particle Deposition by Thermophoresis Under High Temperature Conditions in a Helium-Flow
Authors: Fischer Felix , Andris Andreas , Lippmann Wolfgang , Hurtado Antonio
Published: October 2017

The continuous generation of graphite dust particles in the core of a High Temperature Reactor (HTR) is one of the key challenges of safety during the operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurisation accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions.

On the one hand, the knowledge of the behaviour of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behaviour can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces.

The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5m and a tube diameter of 0.05m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass.

In particular, this test facility offers the possibility to analyse the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a 3D Laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle Image Velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation time are available.

The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for CFD-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of Nuclear Safety.
Scaling and CFD Analyses for a Single Phase Natural Circulation Loop
Authors: Zhang Weizhong , Lin Qian , Si Shengyi
Published: October 2017

Scaling analysis based on scaling laws was performed first for a single-phase natural circulation loop. Two geometrical reduced models were built. Numerical simulations for them using a CFD code named FLUENT were carried out. The calculated heat removal capabilities of the two reduced models show good agreement with the predicted results by the RELAP5/Mod3.2. The reversely deduced heat removal capabilities of the loop prototype from the numerically calculated results agree well with the prediction by the RELAP5/Mod3.2 for full height scaling model.
Numerical Simulation of Core Temperature and Melting Process of IVR Core After a Severe Water Loss Accident
Authors: Liu Yiqun , Zhang Xiaoying , Li Jingya , Wang Biao , Zhan Dekui , Zhang Huiyong , Xu Junying
Published: October 2017

After the occurrence of severe water loss accident in a PWR, the water level in the reactor core would decrease gradually, leading to heat up and melted down of the core, threatening safety of the nuclear power plant and the surrounding environment. In this paper, the 1/4 core of AP1000 PWR was adopted for study, a numerical method has been established to calculate the transient change of temperature and melting process of the core and envelope structure (boarding, basket and RPV) after the severe water loss accident. A two-dimensional conduction model with cylindrical coordinate has been used to simulate heat transfer along the radius and height direction of fuel rods and control rods in fuel assemblies. Heat transfer condition on rod surface considers nucleate boiling for rod surface below the water level, while radiative heat transfer among neighboring rods and natural convection with water vapor was considered for rod surface above the water level. Heat transfer along thickness of envelope structures were modeled with the one-dimensional conduction model. The results show that the maximum temperature of the whole reactor core does not exceed 3000K and AP1000 will not meet the melting of fuel rods with the help of RPV external water chamber cooling. The temperature values of the fuel rods and the control rod show the characteristic distribution of the two regions. At 4904s, the maximum temperature of the rod rises to 2900K, and then stabilize. The temperature of the shell is up to 2000K, the maximum temperature of the basket is to 1260K, the variation of RPV wall temperature is not obvious.
Research of the Minimal Cut Sets Post Processing of the PSA Quantification Engine
Authors: Tang Guofeng , Gao Wei
Published: October 2017

Minimal Cut-sets (MCS) post processing is an important part of the Probabilistic Safety Assessment (PSA) analysis. Two alternative post processing methods have been described in this paper based on traditional and Zero-suppressed Binary Decision Diagram (ZBDD) methods. The first post processing method, named the Quick MCS Post Processing Method, obtains MCS results directly from the cut-sets list using a quick sort method according to a certain set of sorting rules. The alternative post processing method, named the ZBDD Based MCS Post Processing Method, obtains MCS results from the ZBDD structure which encodes the MCS, using the ZBDD algorithm according to the post processing rules. Tests show that both of the two methods can derive the accurate MCS post processing results.
Burnable Poisons Alternative Configurations in AT-FCM PWR Fuel Assembly
Authors: Awan Muhammad Qasim , Cao Liangzhi , Wu Hongchun
Published: October 2017

The Fukushima Daiichi Nuclear Power Plant accident in Japan was one of the severest accidents in history of nuclear power plants. This accident changed the prospect of the Nuclear Engineers all around the globe, regarding safety enhancement of nuclear reactors. Since then many aspects of nuclear reactors regarding the improvements of safety features are under investigation and R&D efforts are underway around the world. Use of innovative fuels in present as well as future reactor designs is one of the major potential areas of these ongoing efforts. Fully Ceramic Micro-Encapsulated (FCM) fuel originally developed for use in high temperature gas cooled nuclear reactors, has proven worth for operating in high temperature environment with high burn-up. Due to its additional fission product barrier in the form of strong SiC layer, it is worth using for application as PWR fuel, thus providing potential benefits related to safety and operational aspects of power plant. However, use of FCM fuel in a PWR also has some operational constraints such as the moderator temperature coefficient (MTC) of reactivity has less negative value and even becomes positive when higher concentration of soluble boron is used for the reactivity control. Thus, use of burnable poison material becomes more important to control the access reactivity throughout the cycle length in such a way the quantity of soluble boron to be used is much lower to prevent the positive MTC value or even soluble boron free operation is possible. In present studies a new candidate designs of PWR fuel assembly of 12×12 square array configuration has been used to study the BP material impact on cycle length. Monte Carlo code MVP-BURN is utilized for the analysis to accurately model the double heterogeneity arising due to TRISO type fuel. Two standard materials i.e. Erbia and Gadolinia are used and different configuration including mixing of BP in matrix material, fuel kernel and in QUADRISO form are analyzed and compared with each other. Impact of the residual poison are also analyzed and additional enrichment required to overcome the impact of residual BP material reactivity are calculated. Different configurations support different BP materials. However, it has been found that with an appropriate combination of both materials and configuration, it is possible to minimize the use of soluble boron. Finally, the recommended assembly configuration is analyzed for MTC value during the entire cycle length, showing sustainability of negative values of MTC for the region of interest. With this kind of arrangements, it is possible to use FCM type fuel for present as well as future generations of the PWRs.
Simulating Fine Particle Deposition in Coolant Channel of Fast Reactor
Authors: Wang Yaoxin , Zhou Tao , Li Bing
Published: October 2017

When fast reactor is operation, because of scouring by high pressure liquid coolant in fuel assembly, there are a lot of products in coolant channel. Using FLUENT software simulate deposition of insoluble particle in fast reactor. Using standard k-ε model predict flow field and turbulence intensity of fluid phase. Using discrete phase model track the trajectory of insoluble particle. The following are simulation results. Fuel cladding deposits lots of insoluble particle, but the concentration of insoluble particle is lower at the central of coolant; Entrance section of the insoluble particle concentration is higher than exit section; Dot deposition of insoluble particle at outlet of fuel cladding will lead to pitting phenomenon, pitting will cause deterioration of heat transfer and destroy the integrity of cladding. In view of deposition law of insoluble particle and characteristic of fuel assembly, mitigation measures of cleaning insoluble particle at fixed time and fixed position are being proposed.
Intelligent Fault Diagnosis Method Based on Operating Parameters in Nuclear Power Plant
Authors: Ke Zhiwu , Hu Xu , Teng Dawei , Tao Mo
Published: October 2017

The safety of mechanical equipment is more important, it directly determines the safety of nuclear power plant operation, and even nuclear safety. So it is necessary to monitor the operating state of NPP system and mechanical equipment in real time by inspecting operating parameters. However, the key technology is real-time fault diagnosis of the mechanical equipment in NPP. Traditional fault diagnosis method based on analytic model is difficult to diagnose relevant and superimposed fault because of model error, disturbance and noise. This paper studies the application of fault diagnosis method based on BP neural network in NPP, and proposes an improved method for neural BP network method. For the feed-water system in the variable load operation process, we select the normal operation, the single feed-water valve fault, feed-water pump and feed-water valve superimposed fault as the analysis objects. One hundred points of data are extracted as BP algorithm training elements in these three processes averagely. The normal and abnormal conditions (including single fault and superimposed fault) can be accurately judged, but the single fault and superimposed failure would produce miscarriage of justice, about 2.4% of the single fault is diagnosed as superimposed fault, the diagnosis time delay is less than 1 second. These results meet the accuracy and real-time requirements. Then we study the application of support vector machine (SVM), which can make up for the deficiency of BP neural network. The results of this paper are useful for the real-time and reliable fault diagnosis of NPP.
Preliminary Analysis of Hydrogen Distribution During Severe Accident Induced by Loss of Coolant Accident
Authors: Wang Di , Cao Xuewu
Published: October 2017

During hypothetical severe accidents in nuclear power plants, a large amount of hydrogen is generated rapidly as a result of Zirconium-Steam reaction and released into the containment. Hydrogen mixes with air and may come into combustion or detonation under proper conditions, which threatens the integrity of containment. Therefore, getting detailed hydrogen flow and distribution in various physical mechanisms is a key issue to resolve the hydrogen risk in containment and compartments. To study local hydrogen distribution in the containment of advanced passive PWR, an analysis model is built by 3-dimensional CFD code. Computational domain is divided by structured grid which contains over 100,000 cells. the shape and surface area of walls and obstacles of steel shell and internal structure, which have great impact on gas flow and heat transfer, are included. Hydrogen distribution in containment simulating with different turbulence models is studied, the result shows that during large amount of hydrogen release stage. In hydrogen distribution result simulating with algebraic model, hydrogen is all gathered in the dome and the peak concentration reaches 17%. When k-ε model is adopted, the peak concentration in the dome is 8%, hydrogen stratification is established in whole large space. Besides, hydrogen distribution near source also shows algebraic model cannot simulate turbulence diffusion in local compartment. It is more reasonable choosing k-ε model to study hydrogen behavior in containment. Based on adopted k-ε model, the effect of steam on hydrogen distribution is investigated. With steam injection, the hydrogen distribution is more homogeneous in upper space and average concentration is lower. In local compartment, due to diffusion enhanced by steam, the hydrogen concentration is higher in the bottom.
Modeling of Two-Phase Boiling Flow and Critical Heat Flux With the NEK-2P CFD Code
Authors: Tentner Adrian , Vegendla Prasad , Tomboulides Ananias , Obabko Aleks , Merzari Elia , Shaver Dillon
Published: October 2017

The paper focuses on the development, implementation, and initial validation of the NEK-2P two-fluid two-phase model. Recent extensions of the Extended Boiling Framework (EBF) models are presented, including the implementation of a four-field two-phase topology which replaces the previous two-field two-phase approach. The paper presents results of recent NEK-2P analyses of several CHF experiments. Good agreement between the simulated wall temperatures and measured data is observed. The simulation results predict well both the dryout location and post dryout wall temperature magnitudes, illustrating the ability of the NEK-2P code and extended EBF models to simulate the CHF phenomena for a range of thermal-hydraulic conditions.
Development of Bubble Departure Diameter Model Based on Force Analysis
Authors: Tian Ye , Huang Wei , Li Pengfei , Cao Simin , Sun Yan
Published: October 2017

Bubble departure diameter has significant effect on bubble dynamics and heat transfer in boiling system, and it is difficult to be measured in a boiling system. Therefore, a method to predict bubble departure diameter is necessary to study of bubble dynamics and heat transfer in boiling system. A new theoretical model based on force analysis is proposed for the prediction of bubble departure diameter in vertical boiling system in this paper. Surface tension force, unsteady drag force, quasi-steady drag force, shear lift force, buoyancy force, hydrodynamic pressure force and contact pressure force are taken into account to build the model. Chen’s experimental data is used to validate the model, the averaged relative deviation between the predict results of the model and the experimental data is less than ±14.8%.
Experimental Study on the Void Fraction of Siphon Breaking Process
Authors: Xing-lin Lei , Shan-fan Huang , Zhong-xiao Guo , Xiao-yu Guo
Published: October 2017

As a safety device to alleviate the loss of reactor coolant, the siphon breaking system is widely used in nuclear power plant. Researchers are very interested in this technique for its “passive” characteristic. Vertical downward air-water two-phase flow is encountered in the siphon breaking process. Previous researches have been more focused on some physical parameters, such as water flow rate, air flow rate, pressure drop and the undershooting height. Void fraction, as a key parameter in multiphase flow, should be studied in the siphon breaking phenomenon. Therefore, a needle-contact capacitance probe is used for flow-phase identification and a single-wire capacitance for obtaining the average value of gas distribution along the straight line. Experimental results show that the flow pattern during the vertical downward air-water two-phase flow is mostly annular flow. With the gas entering the pipeline, void fraction profile against time can be divided into three stages. The slope in the first stage is similar to that in the third. However, the slope slows down in the middle stage. The experimental results also show that the real duration time to break the siphon flow is as short as about 6 s. The void fraction at the end of the siphon breaking process is about 0.38. During this stage, a large amount of gas is sucked into the downcomer and little water is inhaled. The gas phase results in a convergent effect, where the air intake is the direct and fundamental reason of siphon breaking.
Development and Verification of the PWR Lattice Code PANDA
Authors: Zhang Hongbo , Tang Chuntao , Yang Weiyan , Bi Guangwen , Yang Bo
Published: October 2017

Lattice code generates homogenized few-group cross sections for core neutronics code. It is an important component of the nuclear design code system. The development and improvement of lattice codes are always significant topics in reactor physics. The PANDA code is a PWR lattice code developed by Shanghai Nuclear Engineering Research and Design Institute (SNERDI). It starts from the 70-group library, and performs the resonance calculation based on the Spatially Dependent Dancoff Method (SDDM). The 2D heterogeneous transport calculation is performed without any group collapse and cell homogenization by MOC with two-level Coarse Mesh Finite Difference (CMFD) acceleration. Matrix exponential methods are used to solve the Bateman depletion equation. Based on the methodologies, the PANDA code is developed. The verifications on different levels preliminarily demonstrate the ability of the PANDA code.
Research Into the Characteristics of Flow Conditioner in Reactor Coolant Pump Test Circuit
Authors: Chunming Gong , Feng Shen , Ning Bai , Jinghao Li , Likai Fei , Bin Gao , Yanwen Tong , Jinxu Gao
Published: October 2017

During the design process of Reactor Coolant Pump (RCP) test circuit, a flow conditioner is considered to be necessary in view of the fact that the RCP test circuit has a large flow volume and a relatively limited space, and it must meet the need of stabilizing the flow as quickly as possible. This article bases on the demands of the function of the flow conditioner in reactor coolant pump test circuit, utilizes the CFD analysis software to carry out 3-dimensioning modeling of several flow conditioners, and simulates the characteristic of the fluid in circuit under different flow conditioners. Through the analysis of the pressure distribution in the downstream of the flow conditioner, different stabilizing effects are obtained, and the final structure of the flow conditioner is determined considering the real demand of the test circuit.