bluuebunny/crossref_metadata_2025 huggingface dataset:
Design and Development of DeRisk: A Fault Tree Analysis Program Package
Authors: Zhou Zhenxu , Nie Hao , Zhang Qin
Published: October 2018

Fault tree analysis (FTA) has been proven to be a very important tool and has been successfully applied to safety/reliability studies in nuclear, chemical, military, space industries/systems. Hitherto, several useful and popular FTA software/program packages have been developed, like CAFTA+, FAUNET, RiskSpectrum, SAPHIRE, RiskA etc. Minimum Cut Set (MCS) method is the most commonly used traditional FTA method. However, it suffers from low efficiency when solving remarkably large fault trees (FTs). To overcome the shortcomings of the traditional method, several new techniques are proposed such as Binary Decision Diagram (BDD), Zero-suppressed Binary Decision Diagram, (ZBDD) Petri Net (PN), Bayesian Network (BN) and Dynamic Uncertain Causality Graph (DUCG). DUCG is a newly presented Probabilistic Graphic Model to deal with systems with dynamics, uncertainties and logic cycles. DUCG is a good choice to analyze large FTs, in our previous papers, we have proved that any FT can be mapped into a DUCG graph and additional modeling and analytical power can be achieved. DeRisk is a DUCG embedded risk analysis program package written in C# for FTA and is designed as a powerful tool to assist reliability engineers. In this paper, the design schema and the main algorithms of DeRisk are introduced. DeRisk contains five parts: (1) A Graphical User Interface (GUI) Module which interacts with users; (2) A Preprocessing Module which preprocesses FTs (3) An Input Module which allows user to input necessary data by file or by command line; (4) A Calculation Module which offers qualitative/quantitative analysis; (5) An Output Module which outputs the results required by users. Some illustrative examples are used to verify the correctness and effectiveness of DeRisk.
Research About the Uniform Field Breakdown Strength of Helium Gas at High Temperature and Pressure in Millimeter-Scale Gaps
Authors: Qi You , Zhengang Shi , Xingnan Liu , Xunshi Yan , Guojun Yang
Published: October 2018

Helium now commonly serves as the coolant in high temperature gas-cooled reactor (HTGR). Some electrical facilities work in helium gas at certain temperature and pressure, such as the helium blower, magnetic bearings and the electrical penetration assemblies. Because the insulation property of helium is much poorer than that of air, especially at high pressure, the insulation design of these electrical facilities is rather important and challenging. So far, national standards of insulation design is only for the air while there exists almost no industrial standard for helium gas. In addition, not many scientific researches on insulation property of helium have been conducted yet and most efforts were made to deal with liquid helium. In this paper, uniform field breakdown strength in millimeter-scale gaps of helium was studied with a streamer model analysis, at from 15°C to 180°C and 0.01MPa to 7MPa. Coefficients of Paschen law are modified to better fit the experimental data under specific conditions and some analysis or explanation was given. The experimental data and rules obtained will be useful for the design of insulation structure in helium gas and promote the study of HTGR.
Application of Monte Carlo Methods in Reactor Protection System Reliability Research
Authors: Li Duo , Hao Zhaojun , Zhou Shuqiao , Guo Chao
Published: October 2018

Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity.

Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.
A Research on System Error Correction for a High Temperature Hydrogen Detector Based on Neural Network Technique
Authors: Zhenfeng Qi , Yiwang Zhang , Wei Li , Yidan Yuan
Published: October 2018

A mathematical model is established for the High Temperature Hydrogen Detector (HTHD) used in severe accident conditions of nuclear power plants. The system error caused by the temperature difference of the internal wall between the working thermal conductivity cells and the reference conductivity cells is analyzed. Then the back propagation neural network algorithm is introduced to correct the system error. The test results show that BP neural network can effectively suppress this system error, and it has well generalization performance. At the same time, this method can be extended to correct measurement errors caused by other disruptive factors, such as supply voltage fluctuation, velocity variation due to pressure change, and interfering components (e.g. steam).
Research on Auxiliary Bearing Structure With Buffer Shim Based on LS-DYNA for Helium Circulator of HTR-10
Authors: Yang Guojun , Sun Zhe , Liu Xingnan , Shi Zhengang
Published: October 2018

The active magnetic bearing (AMB) is introduced to the helium circulator of the 10 MW high temperature gas-cooled reactor (HTR-10). Auxiliary bearing bears huge instantaneous impact load when the AMB fails to work in the helium circulator. The huge load may cause serious damage to the auxiliary bearing. In order to reduce the impact load and protect the AMB and rotor system, a new type of auxiliary bearing with axial buffer shim is presented in this paper. There are two different structures of auxiliary bearing to be studied in the paper. One is the auxiliary bearing with the buffer shim and another one is without the buffer shim. The finite element method and LS-DYNA software are applied to analyze the structure characteristic of the auxiliary bearing. The rotor dropping trajectory and the stress distribution of the auxiliary bearing are analyzed by comparing two different auxiliary bearing structures during the rotor drop. The stress change of the auxiliary bearing in each impact course is mainly studied, and the law of stress variation in impact course is analyzed. Besides, the stress distribution and deformation of two auxiliary bearing is the focus research when impact force is maximum in the paper. Finally, the research shows the auxiliary bearing structure without the buffer shim can bear the huge impact of the rotor drop, and the addition of buffer shim can also reduce the damage of the rotor drop to the auxiliary bearing structure. These researches’ result provides an important reference for the experiment of rotor drop, and has laid a theoretical foundation for the practical application of this structure.
Estimation of Mitigation Effects of Sodium Nanofluid for SGTR Accidents in SFR
Authors: Ichikawa Kenta , Kanda Hironori , Yoshioka Naoki , Ara Kuniaki , Saito Jun-ichi , Nagai Keiichi
Published: October 2018

Studies on the suppression of the reactivity of sodium itself have been performed on the basis of the concept of suspended nanoparticles in liquid sodium (sodium nanofluid). According to the theoretical and experimental results of studies for sodium nanofluid, velocity and heat of sodium nanofluid reaction with water (sodium nanofluid/water reaction) are lower than those of the pure sodium/water reaction. The analytical model for the peak temperature of a sodium nanofluid/water reaction jet has been developed by the authors in consideration of these suppression effects. In this paper, the prediction method for mitigation effects on damage of adjacent tubes in steam generator tube rupture (SGTR) accidents is developed by applying this analytical model for the peak temperature of the reaction jet. On the assumption that the sodium nanofluid is used for the secondary coolant of sodium-cooled fast reactor (SFR), mitigation effects under the design basis accident (DBA) condition and the design extension condition (DEC) of SGTR are estimated by using this method. The results indicate a clear possibility to reduce the number of damaged tubes and to suppress the pressure generated in SGTR accidents by using sodium nanofluid as the secondary coolant.
Polygonal Virtual Element Spatial Discretisation Methods for the Neutron Diffusion Equation With Applications in Nuclear Reactor Physics
Authors: Ferguson J. , Kópházi J. , Eaton M. D.
Published: October 2018

In this paper the application of the virtual element method (VEM) to the multigroup, neutron diffusion equation will be presented. The VEM is a recently developed Bubnov-Galerkin spatial discretisation method based largely on the mimetic finite difference method (MFD) that preserves the properties of the underlying vector operators. It can discretise elliptic partial differential equations (PDEs) on arbitrary polygonal/polyhedral meshes with arbitrary order and regularity. Deterministic, geometry conforming methods, used to solve multigroup, neutron diffusion, nuclear reactor physics problems have historically used the finite element method (FEM). However, FEM requires high-quality meshes with few highly distorted elements (elements with a poor aspect ratio) or it may may experience convergence problems. The process of creating high quality meshes, even with automated mesh generation algorithms, such as the advancing front and Delaunay methods, is often very time consuming. For these reasons VEM is being studied in this paper as a possible alternative to FEM in the numerical solution of neutron diffusion problems in nuclear reactor physics. A C5G7 UOX pincell problem is presented to demonstrate the application of VEM to mutligroup diffusion problem. The method of manufactured solutions (MMS) is used to determine the order of convergence.
Numerical Investigation on Conjugate Heat Transfer of Supercritical CO2 in Rolling Motion
Authors: Zhao Zhenxing , Liao Mengran , Liu Yong , Xiao Qi , Lao Xingsheng , Wu Jun , Wang Wei
Published: October 2018

The numerical calculations with and without rolling motions were conducted to investigate the effects of ocean environment on flow and heat transfer of supercritical CO2. The AKN k-epsilon model was selected to model the turbulent flow and heat transfer of supercritical fluid. It concludes that the effect of rolling motion on the supercritical CO2 in LPV region is much greater than that on conventional single-phase fluid. The rolling motion can cause the periodic oscillation of the local heat transfer and suppress the nonuniform heat transfer. The secondary flow is induced by both the rolling motion and buoyancy force. The heat transfer is enhanced gradually as the rolling period decreasing or rolling amplitude increasing, but the corresponding pressure drop varies more violently to affect the stability and controllability of the heat exchanger.
Internal Vibration Source Analysis of AMB-Rotor System in HTR-PM Primary Helium Circulator
Authors: Yu Jinpeng , Zhao Lei
Published: October 2018

The active magnetic bearing (AMB) in HTR-PM primary helium circulator (PHC) applies the inductive displacement transducer (IDT) to achieve the closed-loop feedback control. The magnetic anisotropy of the rotor material can be equivalent to structure defect of the rotor and affect the IDT measurement accuracy, leading to internal exciting vibration of the rotor. In this paper, the magnetic field analysis shows that the rotor magnetic anisotropy has effect on the sensor measurement and brings about the displacement measurement error. In the rotor-sensor experiment, the effect of rotor magnetic anisotropy on the IDT is obtained as a curve of magnetic error, which further explains that the rotor magnetic anisotropy will affect the dynamic measurement accuracy of the IDT. With the simulation result, it is observed that the displacement measurement error will lead to the internal exciting force of AMB-rotor system. The force will increase as the rotor speed increases, and bring about high-frequency vibration of the rotor.
In Situ Measurement Technique of Low-Energy β-Contaminated Waste Sorting
Authors: Weihong Yue , Yeqiang Xu
Published: October 2018

Operation and decommissioning of nuclear facilities will produce radioactive waste, and different radionuclides in the waste will bring different hazards to the public and the environment. The waste would be sorted more reasonably by distinguishing different radionuclides. Yet it is still very difficult to measure directly the pure Beta radioactive waste in situ, though in situ Gamma-analytical and Alpha-waste-barrel measurement techniques have become more sophisticated. The aim is to propose a scientific technique to sort the radioactive waste in situ. This study focused on the 90Sr-contaminated material in China Institute of Atomic Energy and optimized the design of the existing solid waste disposal facilities. A novel technique to measure the radioactive waste 90Sr-90Y online was proposed, trying to sort the radioactive waste as optimally as possible to realize further separation of exemption waste. Theoretically, the exemption waste can be further sorted, and it can guide the design of radioactive waste disposal system.
DEM-CFD Study on Fluid Flow Through Radial Layered Composite Packed Beds
Authors: Guo Zehua , Sun Zhongning , Zhang Nan , Ding Ming , Bian Haozhi , Meng Zhaoming
Published: October 2018

The radial porosity generally have a higher value at the container wall than that in the core part. Consequently, the fluid flows are mal-distributed in packed bed with significant bypass flow at the wall, which lowers the convective heat transfer performance inside the packed bed. To overcome this drawback of packed bed, we developed an effective way to construct the radial layered composite packed bed, which can easily realize placing small particle at the near wall region and large spheres in the core region. Therefore, smaller pores forms close to the container wall and larger channels presented in the core part. This could result in a much homogenous radial porosity distribution, which is benefit to restrain the bypass flow near the wall.

In present paper, the packing procedure is simulated by the discrete element method (DEM). Radial layered composite packed bed and traditional packed bed with uniform spheres are compared on radial porosity distribution. By altering the size of spheres ratios in the near wall region different radial layered packings are also generated and compared. Then, the geometries of these packed bed are imported into the computational fluid dynamics (CFD) simulation. The fluid flow inside the packed bed is investigated. It finds that the radial layered packed bed has a lower pressure drop with the ordered packing structure. And a much homogenous fluid flow distribution is obtained than the traditional, which is benefit for the heat removal inside the packed bed. This would be useful for the optimum design of packed bed in industry applications.
COPRA Experiments on Melt Pool Behavior With Eutectic NaNO3-KNO3 Simulant
Authors: Zhou Yukun , Zhang Yapei , Luo Simin , Xu Zhichun , Zhang Luteng , Qiu Suizheng , Su G. H.
Published: October 2018

The COPRA experiments were performed to study the natural convection heat transfer behavior in a large-scale homogeneous melt pool inside the reactor pressure vessel lower plenum. The test section consists of a two-dimensional 1/4 circular slice with an inner radius of 2.2 m. A non-eutectic binary mixture 20%NaNO3-80%KNO3 was selected as melt simulant in the previous tests and the Rayleigh number of the melt pool reached up to 1016. In this paper, the working fluid was a eutectic binary mixture 50%NaNO3-50%KNO3. The melt pool temperature, heat flux distribution and crust thickness were obtained in the experiments with different heating powers. Results from the eutectic molten salt tests can be applied for posttest calculations and comparative analyses.
Research on the Influence of Complex Terrain on Atmospheric Dispersion After Accident
Authors: Wu Nan , Xue Na , Liu Xinjian
Published: October 2018

Emergency condition should be evaluated during the siting of Nuclear Power Plants (NPP). One of the criteria is the capability of emergency evacuation for the purpose of protecting the NPP staff and public. Exposure dose received by evacuees is a key factor to evaluate the effectiveness of emergency evacuation, which can be characterized by the atmospheric dispersion factors over the evacuation paths. Gaussian plume model is usually applied in the atmospheric dispersion factor calculation for NPPs. However, for some NPP sites with complex terrain like mountain and lake which has significant influence on the local flow field and the nuclides transportation, the calculations based on Gaussian model will not be accurate enough for the evaluation of the evacuation capability. In such cases, Computational Fluid Dynamics (CFD) method with fine calculation grids and rich physical models can be used in the atmospheric dispersion modeling of small-scale region, to get much more accurate results. This paper is devoted to the research on the impact of these complex terrain on atmospheric dispersion with CFD method. The local atmospheric dispersion around a NPP site located on an island is modeled, where there is a mountain on the way of evacuation. For the limitation of the geographic barrier, two evacuation paths, which are located closely with small difference in direction, are planned along different sides of the mountain. Utilizing STAR-CCM+ CFD software, local atmospheric dispersion factors are fine simulated under six worse wind directions. Based on the simulation results, the impact of the mountain on the flow field and of the radiological plume on the evacuation paths is analyzed. Results show that due to the blocking effect of the mountain, those two evacuation paths have little probability to be simultaneously covered by the radiological plume. By the choosing of suitable routes during the evacuation, the radiological risk of public is acceptable and emergency evacuation can be carried out effectively. These results also demonstrate that CFD modeling method is effective in the evaluation of local atmospheric dispersion over complex terrain, which can be applied in the analysis of emergency evacuation of NPP sites with complex terrain.
Study on Protection Against Large Commercial Aircraft Crash of HPR1000
Authors: Zhang Xueshuang , Fan Li , Cai Lijian , Liu Qianwen
Published: October 2018

Protection against large commercial aircraft crash is a new design requirement for international advanced nuclear power plants, and gradually becomes one of important advance characteristics. The study follows the principle of balance of safety, economy, and mature achievable engineering technology, conducted by design extension condition and realistic analysis inapplicable to single failure criterion, adopts Anti-Plane Crash (APC) shell protection + redundant features physical isolation to fulfill the safety acceptance criteria, and results in the HPR1000 design scheme withdraw large commercial aircraft crash. Beside the overall analysis on the impact effects of global structure, local structure, vibration and fuel, the study optimizes the design of each outside openings beyond personnel access size in APC shell to limit the mechanical damage of equipment due to local penetration and induced internal fire or explosion by leaking or pouring oil, by the design or construction scheme such as removable walls, protection covers, anti-aircraft gate, manway relocation, concrete second-pouring, and etc. In addition, the study also analyzes the hazard consequence of impact without APC shell protection, especially the influence of the safety structures or equipment damage of other safety buildings (e.g. nuclear auxiliary building) and induced fire spreading or flooding, to ensure the acceptance criteria of integrity of containment and spent fuel pool, or the cooling ability of reactor core and spent fuel pool. This outcome of the study enhances the safety and capability against external extreme hazards of HPR1000, and strengthens the public confidence on nuclear safety and anti-terrorism.
Experimental Research on Energy Release and Fragments Characteristics Under Molten Materials Discharged Into Liquid Sodium
Authors: Hu Liang , Ge Kui , Zhang Yapei , Su G. H. , Tian Wenxi , Qiu Suizheng
Published: October 2018

Hypothetical Core Disruptive Accidents (HCDA) are dominantly concerned during safety assessment and evaluation in Sodium-cooled Fast Reactor (SFR). With molten core materials discharged into liquid sodium, positive reactivity is potentially introduced due to sodium boiling and molten core compaction, which can cause terrible recriticality. The possibility of recriticality and efficient cooling on the relocated debris bed are significantly affected by the fragmentation behavior of molten core in liquid sodium. With few available mechanism models and benchmarks, many investigations have been conducted on the fragmentation characteristics during molten fuel-coolant interaction (MFCI). In the present study, molten copper is used for molten simulant to be discharged into the liquid sodium pool through guiding tube based on a multifunctional experimental facility (COSA). The simulants are heated by electromagnetic induction system in customized ceramic crucible and the molten materials are controlled by magnetic lifting system to be drained through the guiding tube into the bottom liquid sodium pool. Temperature variation and pressure change in the liquid sodium pool are acquired against the energy release during MFCI. Furthermore, the fragments cleaned by water medium are measured and recorded for distribution and morphology analysis. Significant pressure pulses and temperature gradient almost not occur during MFCI and the molten copper is finely fragmented possibly due to hydrodynamic and thermodynamic effects. And the experimental results are helpful to confirm the prediction of fragmentation mechanism and to validate physical model, which can be applied to the development and validation of analysis code.
Classification Optimization for Waste Related Buildings and Structures of NPPs
Authors: Hu Zongwen , Qian Yijie , Fan Li
Published: October 2018

Classification of nuclear power plant waste related buildings and structures is relatively flexible, and research on it is relatively less. In order to better master the classification of waste related buildings and structures and adapt it to the improvement of nuclear power technology and regulations, this paper carries out the function analysis of waste systems and sub-items, and integrates the mature engineering practice experiences and the differences between projects. Finally classification of waste sub-items is analyzed and summarized, such as sectional classification of liquid waste discharge galleries, classification optimization of radioactive waste building. The results of the optimization analysis in this paper can provide sufficient basis and guidance for the classification of nuclear power projects in the future, and improve the economy of nuclear power technology, and shows very good engineering application significance.
Numerical Analysis of Heat Transfer Characteristics of External Cooling Channel of the Lower Head
Authors: Li Dan , Peng Changhong , Guo Yun
Published: October 2018

The heat transfer coefficient of the curved surface reflects the external cooling capacity of the reactor pressure vessel under serious accident. This topic used Ansys fluent software 14.5 as simulation tool, mesh generation of the flow area is carried out after the geometric model of the lower head, and then the heat transfer coefficient of the curved surface is numerically simulated. In this paper, the heat transfer coefficient formulas are fitted according to the laminar and turbulent flow, and the influence of the cavity on the heat transfer coefficient is discussed. Calculation results show that, for single-phase flow both laminar and turbulent flow, the heat transfer coefficients increase with the increase of Re, and increase with the increase of Pr, decrease with the increase of theta; According to the comparison of the heat transfer coefficients under different radii of curved surfaces, reducing the radius is beneficial to increase the heat transfer capability; In subcooled boiling flow, the void fraction is less than 21%, the heat transfer coefficient increases with the increases of the void fraction, the void fraction is conducive to heat transfer; but when the void fraction is higher than 21%, it is opposite.
A Visual Experiment of Single Bubble Growth Processes in a Vertical Rectangular Channel
Authors: Cheng Ning , Guo Yun , Peng Changhong
Published: October 2018

Single bubble growth processes under subcooled boiling condition in a vertical rectangular channel with a gap of 2.8 mm have been visually studied. A high-speed camera was used to observe and record the bubble growth processes at a rate of 6000 frames per second. Four kinds of bubbles with different equivalent radius change trends near the ONB (Onset of Nucleate Boiling) point were observed. The bubble equivalent radius change trends were fitted by traditional empirical formula R(t) = k · tn and found that the value of empirical parameter n was in the range of 0.11 to 0.53 which was smaller than that in literatures, and the value of empirical parameter k had a positive correlation with Ja number. The bubbles generated at the same nucleation point under different heating powers, inlet fluid temperatures and mass flow rates were compared and found that within the experimental range the changes of heating power and inlet fluid temperature had a significant effect on bubble growth rectangular channel which may be used in research nuclear reactor and engineering test reactor. Due to the shape and azimuth of the channel, the bubble behaviors in vertical rectangular channel may be different from those in conventional large-size circular channel or horizontal channel. The investigations reported in the literature generally involved multiple bubbles with complex interactions between the bubbles, and the bubble parameters were usually obtained by averaging the parameters of bubbles generated at different nucleation sites. However, even under the same working condition, bubbles generated by two adjacent nucleation sites may also have big differences. In this experiment, the steady single bubble growth processes in vertical rectangular channel with deionized water as the working fluid were visually investigated. Four kinds of bubbles with different bubble growth curves were observed and the equivalent bubble radius change trends at different heating powers, inlet fluid temperatures and mass flow rates were compared.
A Simulation of Small Break Loss of Coolant Accident in Nuclear Heating Reactor Based on RELAP5
Authors: Lu Meng , Xie Heng
Published: October 2018

Nuclear heating reactor is integrated designed without main pump and safety injection system. The loss of coolant accidents are mainly in the form of small break LOCA. As no safety injection system is designed for coolant makeup, the water volume in the reactor vessel is critical since it determines whether the reactor will be submerged during the whole scenario. Therefore, the study on coolant loss in this pool system is indispensable. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The long term effect in nuclear heating reactor is important. In this paper we investigated the influential factors on SBLOCA scenario and found the long term residual heat removal capacity is decisive in determining the loss of coolant. The residual heat removal capacity should be greater than 2% of reactor thermal power if ensuring the core submerged in the long run.
Uncertainty Analysis of Power Distribution for NESTOR Based on the Double Latin Hypercube Sampling Method
Authors: Liao Hongkuan , Li Qing , Yu Yingrui , Hu Yuying , Wu Lei , Wang Chenlin , Wang Jinyu , Wang Jinghui , Xiao Peng
Published: October 2018

Due to the complexity of the reactor system, many approximations are used in the nuclear design and calculations inevitably. The accuracy of the nuclear design software is closely related to the safety of the reactor design and operation. Besides, improving the accuracy is an effective way to excavating the economy of nuclear power plants. To analyze the uncertainty of power distribution for NESTOR nuclear design software, we suggested an uncertainty analysis method based on the double Latin Hypercube Sampling (LHS) method and the random sampling statistical analysis (RSSA) method, and built an uncertainty analysis process based on LHS. The uncertainty of physical model and the uncertainty of the change of parameters were both taken into consideration with the double samplings, and 3481 core states were generated by the double samplings. Therefore, the uncertainty of power distribution could be directly analyzed through modeling computation, and the uncertainty of radial power distribution was achieved as ±3.856% under the condition of 95% confidence coefficient and 95% probability. Meanwhile, according to deviation transmission idea, we obtained the uncertainty of power distribution from physical models and the change of parameters based on the measured method. The result shows that the accuracy using the double sampling method is nearly the same to which achieved by the deviation transmission idea, and more conservative.
Turbulent Transverse Plane PIV Measurements on a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle
Authors: Goth Nolan E. , Jones Philip , Nguyen Thien D. , Vaghetto Rodolfo , Hassan Yassin
Published: October 2018

This study recaps the experimental effort to characterize a transverse plane flow field in a tightly-packed, 61-pin, wire-wrapped, hexagonal fuel bundle prototypical for a sodium fast reactor. The motive was to produce high spatiotemporal experimental data for computational fluid dynamics turbulence model validation. The matched-index-of-refraction and laser-based optical measurement techniques were utilized on an isothermal experimental flow facility. Measurements were performed on a transverse plane perpendicular to the axial flow. Fluid flow in the three types of subchannels (corner, exterior, and interior) were quantified. All measurements have been performed at a bundle-averaged Reynolds number of approximately 10,400. Results include flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, and Reynolds stress. Of interest was the flow behavior around the restriction caused by the wire spacer and hexagonal duct wall, where recirculation regions formed. Regions of maximum and minimum momentum transfer coincided with regions of maximum and minimum fluctuations. These regions highlight locations of maximum and minimum cooling of the fuel pins. The experimental data will be used to benchmark computational fluid dynamics simulations of the sodium fast reactor fuel bundle using Reynolds-averaged Navier Stokes and large-eddy simulation turbulence modeling methods.
Development and Application of Level 3 PSA for the UK ABWR Generic Design Assessment
Authors: Grindon Elizabeth , Harman Neil , Niculae Carmen , Ang Ming Leang , Iwanami Hironobu , Hashimoto Tomoharu
Published: October 2018

A full scope Probabilistic Safety Assessment (PSA) was provided as an integral part of the safety case for the UK ABWR Generic Design Assessment (GDA) and this included a Level 3 PSA. The main objectives of the Level 3 PSA for GDA are to provide a demonstration of the compliance, for a single unit UK ABWR, with numerical risk targets defined in the UK Safety Assessment Principles and to support the ALARP assessment. This paper includes:

• An overview of the methodology, PSA model development and illustration of some results.

• A summary of the key assumptions made during the model development for the GDA phase of the project. Compliance with the numerical risk targets has been investigated through assessments against the individual off site risk from the facility (Target 7), facility dose bands (Target 8) and off site societal risk (Target 9).

• Some conclusions of a peer review against the draft ASME/ANS standard for trial use (ASME/ANS RA-S-1.3, Feb 2016 for Level 3 Probabilistic Risk Assessment (PRA)), which was a key aspect of this study.
Transient Boiling and Cross Flow in 5×5 Rod Bundle With Rapid Heating
Authors: Takiguchi Hiroki , Furuya Masahiro , Arai Takahiro , Shirakawa Kenetsu
Published: October 2018

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped.

This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 × 5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40°C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.
Experimental Study on Spray Pattern of Pressure-Swirl Nozzle in Reactor Containment
Authors: Bian Jiawei , Zhang Dalin , Sun Rulei , Wu Yingwei , Tian Wenxi , Su G. H. , Qiu Suizheng
Published: October 2018

Spraying system plays an important role in the safety of PWR. To ensure homogeneous spraying of the containment, the layout of nozzles on the spray header was taken into consideration in design. In this paper, an experimental study was conducted to obtain spray characteristics data, including spray cone angle and 2-D spray flux distribution for the purpose of achieving optimal design of the spraying system.

According to the specialty of the spray field involved, a testing loop with four pressure-swirl nozzles was established for the study. Spray cone angles were obtained by photograph method. The volume flux distribution was measured by collecting the spray droplet along the cross-section diameters. Based on the experimental data, typical spray flux distributions were obtained. The flux distribution results were used to build 3-D coverage models. Then these models were used to calculate the overall spray coverage in the containment. The present work introduces the experimental study of spray behavior of a typical pressure-swirl nozzle in containment and the method to evaluate spray coverage through building 3-D spray flux distribution models. The work is expected to be helpful for the optimization design of spraying systems.
Eulerian Two-Fluid Model for Aerosol Removal in Filtered Containment Venting Scrubbers
Authors: Kim Ji-Su , Park Jong Woon , Kim Minkyung
Published: October 2018

Eulerian steady one-dimensional two-fluid mass and momentum conservation equations for gas and water droplet coupled with the particle removal rate are formulated. The major flow variables are phasic velocities and pressure and they are obtained by integrating the four conservation equations and one particle removal rate along the horizontal throats of the venturi scrubbers by using the forth-order Runge-Kutta method. Popular correlations are used for the impaction parameter, the gas-to-water droplet drag coefficient and the droplet size. All the equations are integrated into one computer code named REMAERO based on the MATLAB. It is found that most particles are removed at the throat entrance as already known from the previous experiments and this is due to the particle removal rate proportional to the particle concentration. Particle removal efficiencies from the present method shows better agreement with existing experimental data than previous one-dimensional models.
Development and Application of a 2D/1D Fusion Code With Leakage Reconstruction Method
Authors: Liang Liang , Liu Zhouyu , Wu Hongchun , Wang Sheng , Zhang Qian , Zhao Qiang
Published: October 2018

The 2D/1D fusion method (2D/1D method) is becoming a popular transport method for whole-core calculations, which reduces the group condense and assembly homogenization approximations in the conventional two-step reactor physics calculations. In most 2D/1D codes, a pin is chosen as a 1D calculation domain, which assumes that the axial leakage of the pin is flat on top/bottom surfaces. Similar to the axial leakage, the radial leakage of every 2D plane also introduces several approximations along axial direction for the 1D calculation. In this paper a 2D/1D fusion code is developed, while a leakage reconstruction method is proposed and applied. In this 2D/1D fusion code, MOC is applied to the radial 2D calculation and the Sn diamond difference method is used for the axial 1D calculation. Numerical results indicate that the 2D/1D fusion code developed in this paper is precise in three-dimensional transport calculation and show the performance of this leakage reconstruction method especially when the leakage term is significant.
Development and Verification of a New Nuclear Data Processing System NECP-Atlas
Authors: Xu Jialong , Zu Tiejun , Cao Liangzhi , Wu Hongchun
Published: October 2018

To process the evaluated nuclear data file (ENDF) libraries and generate the cross section data library for neutronics calculations, a new nuclear data processing system NECP-Atlas was developed by Nuclear Engineering Computational Physics Lab. of Xi'an Jiaotong University. Meanwhile, some flaws of the current widely used nuclear data processing systems were made up.

Some new methods and techniques were proposed and integrated into NECP-Atlas. NECP-Atlas could process ENDF and generate point-wise evaluated nuclear data file (PENDF) and the multigroup cross section data library in WIMS-D format. Verification of NECP-Atlas was carried out by comparing the keff values for WLUP benchmark cases and benchmark experiments in the ICSBEP handbook using cross section data libraries processed by NECP-Atlas with those by NJOY2016. The results showed that NECP-Atlas processes the ENDF correctly and generates more reliable cross section data libraries.
Structure Innovation and Thermal Hydraulics Analysis for Cold Trap of CEFR
Authors: Zu-guo Chen , Yi-jun Xu , Wen-jing Qi
Published: October 2018

As one of the significant equipment in CEFR operation system, cold trap plays an important role in purifying the coolant sodium and ensuring the safe and steady operation of the reactor. After several years’ of operation, the cold trap of CEFR can’t work very well as it has been designed previously, which results in the low impurity aggradation efficiency, being easily blocked and frequent replacement. Analyzing the structure of the current cold trap, there are two possible reasons causing the problems mentioned above. The first one is that the sodium flow path is improper in the cold trap resulting in the blockage in specific area. The second one is that the density and layout of wire mesh didn’t take the zoning scheme, which would make the precipitation of impurities become easier in exterior area and the flow path is thus blocked.

Based on the two reasons, a modified scheme is proposed from the view of structure and hydraulics to improve the impurities precipitation mode so that the lifetime of the cold trap can be prolonged. The modified scheme is to block about 1/3 orifices on interior cylinder wall in wire mesh zone and to open two rows of diversion holes on the upper support plate of wire mesh so that the velocity distribution can be improved. Meanwhile, in order to make impurities precipitate from the bottom of the cold trap, the wire mesh is arranged to be in upper, middle and lower areas using different wire mesh density. By 3-D modeling and numerical analysis and then comparing the results to that of current cold trap of CEFR, it can be concluded that the velocity distribution in modified scheme is more reasonable and the new layout of the wire mesh can meet the impurities precipitation requirements better. Further comparative analysis with the results under the actual operational condition will be carried on after the completion of the manufacture for modified cold trap.
Analysis on Flow Behavior in the Plenum of RPV of PWR
Authors: Huang Lei , Hao Lu-lu , Chen Hong , Xue Jun-feng , Tong Li-li
Published: October 2018

The flow distribution at core inlet plays a vital part on hydraulic design of pressurized water reactor. Nonuniform coolant flow distribution at core inlet is caused by many factors, among which the behavior of flow in the lower plenum is the most direct cause. Therefore, a further research on the flow behavior of coolant in the lower plenum is very important and necessary. However, the flow behavior is dominated by the variation of the flow environment related to the complex structure and the condition of upstream uniformity. Using CFD methods, the flow field in the pressure vessel under the uniform flux operation condition of three-loops is simulated in this paper. The standard k–ε turbulence model and upwind solver scheme are selected. Pressure and velocity along with the flow direction are investigated. The variation trend of flow characteristics is discussed by analysis on streamlines at different locations ranging from inlet pipe to lower core plate, which provides evidence for the formation of swirling eddies in the lower plenum and uneven flow flux at lower core plate. In order to understand the formation process of eddies of different sizes in the lower plenum, the velocity fields in the downcomer and lower plenum under the conditions of different inlet velocity are analyzed. Furthermore, the effects of key structures on the formation of swirl are also presented. The results show that the value of velocity flowing into the lower plenum is an important factor on the range of vertical reflux effect, also slowing down because of the resistance of the vortex suppression plate. And the value of angular momentum of swirls in the lower plenum is mainly determined by the inlet angle of converging streams flowing from the downcomer, which is caused by dispersed flow along the cylinder wall flowing from two cold legs, related with the inlet conditions.
A Methodology to Determine SMR Build Schedule and the Impact of Modularisation
Authors: Lloyd Clara A. , Roulstone Anthony
Published: October 2018

Light-water cooled Small Modular Reactors (SMRs) are a potential game-changing technology for energy supply. The potential benefits of SMRs are however conditional on solving the key standardisation and construction issues that have troubled large reactor (LR) projects, which have in turn led to high build costs and long project durations.

Initiatives to determine the build schedule of SMRs are hindered by a lack of SMR construction experience and related data. The methodology used in this paper, to deal with the lack of SMR-specific data, draws conclusions about SMRs based on data from actual large pressurised water reactor (PWR) construction experience.

It is expected that SMR build schedules can be greatly reduced because of the smaller physical size of structures, fewer components, and other size-related features. However, the construction work space will be more constrained, which could negatively impact build durations. As a result, simple geometric scaling and reduction arguments cannot necessarily be applied to SMR schedules. This paper defines the key areas in which SMR construction differs from LRs, such as smaller geometries as well as modularised and standardised build processes, and describes how these differences might be expected to impact build duration quantitatively.

The model developed in this paper presents an approach to determining SMR build schedule durations for a range of reactor sizes. It starts with an LR build schedule based on real data from the UK’s only PWR, Sizewell B. The available data are used to establish a reference point for a non-modular, stick-built SMR schedule. This scheduling approach assumes that, for each major element, part of the time spent on fabrication and installation tasks will vary with reactor size while the remaining fraction will remain constant regardless of reactor size (e.g. due to quality assurance and commissioning tasks). The accuracy of the model generated here is assessed against available construction data and models from a range of actual reactor build projects.

The objective of this work is to consider how modularisation can reduce build schedule of SMRs of varying size, by employing modular design and construction principles to both remove tasks that are of long duration from the critical path and to improve construction productivity. Mechanisms by which modularisation reduces build schedule are investigated. Build reduction scenarios are presented based on analysis and subsequent modularisation of the SMR critical path and are compared with other related analyses.
Validation of a Code and Effect of Turbulence Model on Predicting Thermal Stratification Phenomena in the Upper Plenum of SFR
Authors: Wang Shibao , Zhang Dalin , Wang Chenglong , Song Ping , Chen Jing , Qiu Suizheng , Su Guanghui
Published: October 2018

Thermal stratification phenomena occurring in the upper plenum during a scram transient have an important influence on the structural integrity and the passive safety of sodium-cooled fast breeder reactor (SFR). A two-dimensional thermal-hydraulic analysis code was developed under cylindrical coordinate based on conservation laws of mass, momentum and energy. Block-structured grids were generated to resolve the problems with complicated geometric properties. A second-order scheme based on midpoint rule was applied for the discretization of convection and diffusion terms. Two RANS-type turbulent models, i.e. the standard k–ε model (SKE) and the realizable k–ε model (RKE), are available in this code. A sodium test with scaled model, characterized by large aspect ratio, of a Japanese prototype SFR was used for the validation, mainly from the viewpoints of vertical temperature profiles and rising characteristics of the stratification interface. Results showed that this code could reproduce overall basic behaviors of thermal stratification. The sodium with higher temperature stayed largely stagnant in the upper region under buoyancy effect. Due to the high heat conductivity of sodium, momentum transportation made its leading function. Thus, the RKE model which accounts for the mean deformation rate gave better outcomes than the SKE model.