bluuebunny/crossref_metadata_2025 huggingface dataset:
Analytical Study of Supercritical Water Flow in Two Heated Parallel Channels With Wall Heat Effects
Authors: Ghadge Dhanashree S. , Chatoorgoon Vijay
Published: October 2018

One dimensional (1D) numerical simulations have been carried out of heated parallel channel experiments using the linear frequency domain approach to investigate the effect of wall heat on the system’s stability for supercritical water in vertical heated channels. Oscillatory instability boundaries of the experimental cases with two different wall thicknesses and uniform power distribution are produced, assessed and discussed. These new numerical results with wall heat are compared with experimental data and previous numerical results without wall heat. There is improved agreement on the stability boundary predictions when the wall heat storage effect is included. However, the effect of wall heat storage was small for these experiments.
Advances in FEMN: The Code for Nuclear Noise Analysis Based on ICEM-CFD
Authors: Yuan Baoxin , Zeng Herong , Yang Wankui , Zhang Songbao
Published: October 2018

The finite element method based on unstructured mesh has good geometry adaptability, it has been used to solve reactor physics problems, manual description of geometric modeling and meshing makes the current finite element code very complicated, it greatly restricts the application of this method in the numerical calculation of reactor physics. Using the CAD pre-processing software ICEM-CFD, three dimensional geometry is divided into tetrahedral or hexahedral meshes, two dimensional geometry is divided into triangular or quadrilateral meshes, the main code of neutron calculation for nuclear noise analysis based on finite element method is developed. The steady state parameters are calculated and tested through benchmark problem, the test results show that the code has the corresponding computing capabilities. Finally, the neutron noise spectrum is calculated for the 3D PWR benchmark problem published by IAEA, and the noise distribution under given frequency is given.
Energy Calibration of Scintillator Detectors in Different Neutron Diagnostic System on Tokamak
Authors: Cui Zhiqiang
Published: October 2018

The purpose of tokamak plasma diagnostics is to provide the necessary parameters for device protection, operation, and maintenance. It can also supply parameters for fusion physics research. As one of the main ways to diagnose nuclear fusion plasma, neutron diagnosis focuses on the detection of neutrons, produced by the D-D and D-T fusion reactions, to obtain the physical information of internal plasma. Neutron measurements are widely performed on tokamak to provide the essential information on the neutron yield rate of the plasma that is related to fusion power. Since neutron has no electric charge, neutron can’t be ionized directly by the interaction of electrons in the detection material. The interactions between neutron and nuclei, such as nuclear reaction and nuclear recoil, are used to detect neutrons. According to the front sensitive materials, neutron detectors can be divided into gas detectors, scintillator detectors, semiconductor detectors, ionization chambers and so on. Since the magnetic field surrounding Tokamak can have a magnificent influence on the performance of photo-electronic multiplier tubes (PMTs), it is necessary to employ magnetic shielding in designing detectors, thus guaranteeing the proper operation of detectors within a strong magnetic field. Although the PMTs are equipped with magnetic shielding materials by manufacturers, they can only resist the influence of geomagnetic field. Besides the magnetic shielding and neutron/gamma shielding, neutron detectors should be calibrated before used on the tokamak. Nine similar detectors were assembled and calibrated in this paper. The basic idea of processing calibration data is that we should adjust the resolution and the light response function in order to make experiment spectrum and simulation spectrum fit on the recoil proton edge. A special explication is given to the data processing of neutron calibration, followed by an analysis of its resulting light response function and by comparison with PTB’s results.
A Space-Time Parallel Method to Solve Space-Dependent Neutron Kinetics Equations in Hexagonal-Z Geometry
Authors: Zhang Zhizhu , Cai Yun , Peng Xingjie , Li Qing
Published: October 2018

Neutron kinetics plays an important role in reactor safety and analysis. The backward Euler method is the most widely used time integration method in the calculation of space-dependent nuclear reactor kinetics. Diagonally Implicit Runge-Kutta (DIRK) method owns high accuracy and excellent stability and it could be applied to the neutron kinetics for hexagonal-z geometry application. As solving the neutron kinetics equations is very time-consuming and the number of available cores continues to increase with parallel architectures evolving, parallel algorithms need to be designed to utilize the available resources effectively. However, it is difficult to parallel in time axis since the later moment is strongly dependent on the previous moment. In this paper, the Parareal method which is a time parallel method and implemented by MPI in the processor level is studied in the hexagonal-z geometry with the help of DIRK method. In order to make good use of the parallelism, a parallel strategy in the space direction is also used. In the coarse nodal method, many same operations are finished in the nodes and these operations could be parallel by OpenMP in the thread level since they are independent. Several transient cases are used to validate this method. The results show that the Parareal method gets a fast-convergent speed such as only 2∼3 iterations are needed to convergent. This space-time parallel method could reduce the cost time compared to the sequential method.
Design Study of Beam Window for Accelerator-Driven System With Subcriticality Adjustment Rod
Authors: Sugawara Takanori , Eguchi Yuta , Tsujimoto Kazufumi , Obayashi Hironari , Iwamoto Hiroki , Matsuda Hiroki
Published: October 2018

Engineering feasibility of the beam window is one of the design issues in the accelerator-driven system (ADS). This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the required proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA.

Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the following design; the hemisphere shape, the outer radius = 180 mm, the thickness at the top of the beam window = 1.5 mm, and the factor of safety for the buckling = 3.8, was presented. The buckling pressure was almost same as the previous one and more feasible beam window concept was presented through this study.
Condition Fault Tree: An Extension of Traditional Fault Tree to Handle Uncertainty
Authors: Zhou Zhenxu , Zhang Qin
Published: October 2018

Fault Tree Analysis (FTA) is a powerful and well-established tool, widely-used to evaluate system reliability. The logical connections between faults and causes in Fault Trees (FT) are assumed to be deterministic and are represented graphically via logical gates (such as AND gate, OR gate, NOT gate, etc.). However, sometimes the causalities can be uncertain. Considering that some of the causal relationships in FTs may be uncertain or non-deterministic, we propose a new model to represent the uncertainties, so called as Condition Fault Tree (CFT). We extend the traditional FTA by introducing a new parameter U, which illustrates the random mechanism of how parent event can cause child event. The probability of U (which is denoted by u = Pr{U}), is used to measure the uncertainty between parent event and child event. By introducing rules of parameter U in CFT, we explore its properties and corollaries. We also introduce a methodology to simplify CFTs based on Contraction, Elimination and Extraction rules. With the simplification rules, the structure of CFT can be simplified and the size of CFT can be significantly reduced. Since CFT is an extension of traditional FT, a qualitative analysis method and a quantitative method are introduced. For qualitative analysis, one can simplify a given CFT into the simplest form with the aforementioned rules, properties, and corollaries. With the simplest form of CFT, one can then get the Minimum Cut Sets with uncertainties, as an extension of Minimum Cut Sets. For quantitative analysis, exact calculation methods based on Inclusion-Exclusion and Disjoint-Sum-of-Product are proposed. Some examples are used to illustrate how CFT works.
Interface Tracking Simulations of Two-Phase Flow Utilizing Adaptive Meshing Capabilities
Authors: Cambareri Joseph J. , Bolotnov Igor A.
Published: October 2018

Due to the increase of computing efficiency and power, full-resolution two-phase flow simulations have become a practical research tool for model development and analysis of reactor flows. The expansion of state-of-the-art high performance computing (HPC) facilities allows for the use of direct numerical simulation (DNS) coupled with Interface Tracking Methods (ITM) to perform full resolution simulations. Given adequate spatial and temporal resolution, DNS can resolve all relevant turbulent scales, allowing for the extraction of high quality and detailed turbulent and two-phase flow numerical data for use in model development. While larger scale bubbly flow DNS are becoming ever more affordable, it is still computationally expensive due to the requirements of the spatial discretization. This presents the largest obstacle for future applications of DNS.

For this reason, mesh adaptation techniques are sought after to reduce the computational expense of bubbly flow simulations in complex geometries. By fully resolving only the areas of specific interest, the computational costs of DNS can be reduced. Grid refinement can be based on the location of the interface between the two phases, area of greatest turbulent intensity, averaged bulk fluid velocity data, or the prediction of bubble movement. Coupled with an advanced bubble tracking algorithm, the path of individual bubbles moving through the computational domain can be predicted, and the computational mesh refined within the path area. This refinement can create tracks of greater resolution for the bubbles to move through in the domain, while keeping the bulk resolution of the mesh coarser. Through these means, the overall cost of the simulation is reduced, while high quality numerical data is still obtainable. This work outlines the enhancement of existing mesh adaptation algorithms to implement the bubble tracking refinement, and its practical applications to full resolution two-phase flow simulations.
Development of Neutronics and Thermal-Hydraulics Coupled Code for Accelerator Driven Subcritical Systems
Authors: Luo Run , Wang Pengfei , Wei Xinyu , Revankar Shripad T. , Zhao Fuyu
Published: October 2018

A new neutronics and thermal-hydraulics coupled code named ARTAP is developed to analyze the steady-state and transient characteristics of accelerator driven subcritical system (ADS) in this paper. Monte Carlo simulations for spallation neutron source and deterministic calculations for the subcritical core are performed in the steady-state analysis module. The ADS core is divided into a number of nodes both along the reactor axis and the fuel pellet radius for the calculation of spatial temperature distributions. The power iteration method is adopted to solve the coupled neutronics and thermal-hydraulics problems. The transient analysis module consists of space-time neutron kinetics model and thermal-hydraulics dynamic model, which is calculated by using numerical differentiation formulas (NDFs) method. The new code is verified by comparing its predictions for both the steady-state and transient cases of the OECD/NEA ADS benchmark. Results of numerical simulations indicate that ARTAP is reliable and efficient to be applied for the ADS analysis.
Two-Phase Flow Regime Identification Using Fluctuating Force Signals Under Machine Learning Techniques
Authors: Saito Yuta , Torisaki Shuhei , Miwa Shuichiro
Published: October 2018

For rational design of industrial machineries such as nuclear power plants and heat exchanging devices, understanding of the two-phase flow regime is crucial. In this study, a new method of flow regime identification using the two-phase fluctuating force signals is proposed. Unlike the existing methodologies to measure two-phase flow parameters, the advantageous feature of utilizing the fluctuating force signal is that the measurement can be conducted under completely intrusive environment. Experiments were conducted using the tri-axial force transducers installed at the 90 degrees pipe bend of the vertical upward flow. For signal classification, machine learning techniques were utilized to identify flow regime, and four types flow regimes, namely, bubbly, slug, churn-turbulent, and annular flows were considered. From the obtained fluctuating force database, the features that characterize the signal were selected in the time and the frequency domain. In the current study, three types of machine learning algorithms such as the artificial neural network (ANN), support vector machine (SVM), and decision tree were examined and results obtained by each learning technique was compared.
A Simplified Force-Balance Model to Predict Bubble Departure Diameter in Horizontal Flow Boiling
Authors: Du Jingyu , Zhao Chenru , Bo Hanliang , Zhou Yujia
Published: October 2018

In this paper, a simplified force-balance model in horizontal flow boiling is proposed by introducing several dominant dimensionless parameters. Forces acting on a bubble attached in the direction parallel to the heating surface are analyzed completely, which include quasi-steady drag force, surface tension force and growth force in x-direction. When the force balance is broken in the x-direction before the y-direction, bubble will departure from the nucleation site immediately. Based on the force-balance analysis, bubble departure diameter is formulated to be a function of Jakob number, Reynolds number and Prandtl number. By analyzing the effects of dimensionless parameters, it is found higher Jakob number indicates larger bubble departure diameter, while the increase of Reynolds number will reduce the departure diameter. Besides, the simplified model is able to enlarge application of force-balance model under various dimensionless parameters and make the calculation of bubble departure diameter much easier. Finally, comparing with the experimental data reported in literature, the error between simplified force-balance model and experimental is about 25%.
A Dimensional Analysis of Ex-Vessel Steam Explosion
Authors: Wei Li , Qiang Guo , Yidan Yuan
Published: October 2018

This paper presents a dimensional analysis for ex-vessel steam explosion occurring in a severe accident of nuclear power plants. Some dimensionless parameters were obtained from this analysis through Buckingham’s PI theorem. These dimensionless parameters are not only helpful to design experimental facilities and testing parameters, but also useful to categorize the experimental findings into uniform forms which facilitate data mining for in-depth understanding and generalization.
Development of Experimental Facility to Study Channel Disassembly Behaviour for Indian PHW Reactor During Heatup
Authors: Sahoo Pradeep K. , Singh Ankit
Published: October 2018

No experimental investigation has been carried out to understand the progression of channel disassembly in IPHWR. Hence, it is planned to study the deformations aspects through experiments with a scaled-down (1:3) fuel channel of 220MWe reactor. Depending upon the pressure inside the channel and transient past, the accident initiation and disassembly can be can be classified into two class. In the first case, high pressure inside channel leads PT ballooned and fully contacted CT before channel deformed as one composite unit. While in second case, low inside pressure, first leads PT to contact CT and then coupled deformation of the channel progresses. Latter is part of our current experimental investigation.

This paper is focused on discussion about various aspects considered while developing the experimental facility. In this case, experiments are grouped under two broad categories, one is the study under inert environment to check channel deformation due to material flow at higher temperature and other one is under oxidation environment, which would be similar to actual condition exist in normal reactor core under accident scenario. Initially experiments will be carried out on single channels, later it will be extended up to three channels arranged one above other. A simulation study is also carried out to check the deformation profile under inert conditions. Simulation is done using coupled field element technology of commercial software ANSYS v17.2, to get a preliminary insight into structural and thermal coupled response on channel deformation.
Buildup and Decay Analysis of Corrosion Products Activity in Primary Coolant Loop of AP-1000
Authors: Mahmood Fiaz , Hu Huasi , Cao Liangzhi
Published: October 2018

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.
Comparison of Drift-Flux Models for Void Fraction Prediction in Sub-Channel of Vertical Rod Bundles
Authors: Ren Quanyao , Pan Liangming , Zhou Wenxiong , Ye Tingpu , Liu Hang , Li Songsong
Published: October 2018

In order to simulate the transfer of mass, momentum and energy in the gas-liquid two-phase flow system, tremendous work focused on the phenomenon, mechanisms and models for two-phase flow in different channels, such as circular pipe, rectangular channel, rod bundle and annulus. Drift-flux model is one of the widely used models for its simplicity and good accuracy, especially for the reactor safety analysis codes (RELAP5 and TRAC et al.) and sub-channel analysis code (COBRA, SILFEED and NASCA et al.). Most of the adopted drift-flux models in these codes were developed based on the void fraction measured in pipe and annulus, which were different with the actual nuclear reactor. Although some drift-flux models were developed for rod bundles, they were based on the void fraction on the whole cross-section not in subchannel in rod bundles due to the lack of effective measuring methods. A novel sub-channel impedance void meter (SCIVM) has been developed to measure the void fraction in sub-channel of 5 × 5 rod bundles, which is adopted to evaluate these existing drift-flux models for rod bundles. By comparison, the values of drift-flux parameters have large differences among different correlations, which are suggested to be reconsidered. Based on the experimental data and physical laws, Lellouche-Zolotar and Chexal-Lellouche correlations show a better performance for drift velocity. If the predicting error of void fraction is the only concerned parameter, Chen-Liu, Ishizuka-Inoue and Chexal-Lellouche correlations are recommended for averaged relative error less than 30%. More experiments are suggested to focus on the distribution parameter and drift velocity through their definition.
Pressure Drop Experiments of Liquid Sodium Flowing in a 7-Rod Bundle
Authors: Hou Yandong , Wang Liu , Wu Yingwei , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: October 2018

Pressure drop experiments was conducted for liquid sodium in an electrically heated 7-rod bundle. The electrically heated 7-rod bundle was placed in a hexagonal tube. In the experiment, the heat flux ranges from 0∼300 kw · m−2, mass velocity from 40∼450 kg · m−2 · s−1, system pressure from 10∼200 KPa and the average temperature of liquid sodium from 350∼650°C. The effects of the heat flux, system pressure and the average temperature of liquid sodium on the pressure drop was in-depth analyzed. A new correlation for pressure drop was developed based on the experimental data of liquid sodium in a 7-rod bundle.
Calculation of Core Damage Frequency Caused by Main Control Board Fire in the Main Control Room for Small Modular Reactors
Authors: Wang Wanhong , Peng Changhong , Guo Yun
Published: October 2018

Fire accidents have had an adverse impact on nuclear safety in nuclear power plants. Fire probabilistic safety assessment, as one of the most important methods for safety analysis, has played an essential role in fire risk in nuclear power plants. Main control rooms normally contain many cabinets equipped with all kinds of cables. Once cables there get burned, if not tackled immediately in an appropriate way, they would possibly cause severe consequences. In order to assess the fire risk in the referenced main control room (MCR) for a small modular reactor, this study used the software, CFAST, to simulate fire scenarios caused by the main control board in the MCR, and use the software, Risk Spectrum, to calculate the Core Damage Frequency (CDF). The calculation results show that CDF caused by the main control board (MCB) is 6.112E−09/year.
First-Principles Studies of Diffusion Behaviors of Tritium in HTR-PM Materials: From Framework to Preliminary Result
Authors: Fang Chao , Wang WenYi , Chen HongYu , Li Chuan
Published: October 2018

High Temperature Reactor-Pebblebed Modules (HTR-PM) is a typical high-temperature gas cooled reactor (HTGR) [1]. Tritium is one of the most important radionuclides in reactors owing to its very harmful β-radiation and long half-life. In the HTR-PM, Silicon carbide (SiC) is the main barrier of triisotropic (TRISO) particles to prevent the diffusion of Tritium into the primary circuit [2]. When Tritium into the primary circuit and circulate to Steam Generator (SG), the Incoloy800H alloy is another important material to prevent the diffusion of Tritium into the secondary circuit [3]. When analyzing the source term of Tritium in HTR-PM primary and secondary circuit, it is important and necessary to know the diffusion behavior of Tritium in SiC and Incoloy800H and furthermore, the detail mechanism of diffusion is also essential, which could not be obtained from traditional phenomenological analysis and conservative estimation. In order to solve this challenge, a framework with ab-initio methods is established. In this paper, the detail theory of ab-initio theory and the actual usage in the calculation of the diffusion path, barrier energy are given firstly. And then, the most physical path and the minimum energy barrier will be determined, which can be considered as the diffusion activation energy. The calculated results of activation energy of Tritium in SiC and Incoloy800H are 0.442eV and 0.757eV respectively. Furthermore, the theoretical results are compared with the experimental data, and it is found that both are in agreement with each other. These results are very helpful for understanding the diffusion behaviors of Tritium in HTR-PM materials and can be used to guide the tritium source term analysis in HTR-PM, which are first studied from a micro perspective.
Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique
Authors: Li Xing , Tan Sichao , Mi Zhengpeng , Qi Peiyao , Huang Yunlong
Published: October 2018

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.
Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys
Authors: Shi Lin , Chen Liutao , Xu Yang , Gao Changyuan , Tan Jun , Deng Yongjun
Published: October 2018

Two kinds of new cladding material for pressurized water reactor were developed by China Nuclear Power Research and Technology Institute (CNPRI) which named CZ1 and CZ2. The cladding tubes of CZ alloys used in the study were fabricated by different final annealing temperature in the range of 450 °C to 600 °C.In order to investigate effect of final annealing temperature on creep property of CZ alloys. The axial creep tests were carried out on specimens of CZ-SRA and CZ-RXA At the temperature of 375 °C with different applied stress levels for 1000 hours. The experimental results showed that the axial creep resistance of CZ-RXA is superior to CZ-SRA at lower stress and is inferior to CZ-SRA at higher stress. The mechanism was discussed.
Study on the Prediction of DNB-Type Critical Heat Flux in Rod Bundle Under Motion Conditions
Authors: Huang Siyang , Wang Xiaoyan , Tian Wenxi , Chen Ronghua , Wu Junmei , Su G. H. , Qiu Suizheng
Published: October 2018

In the nuclear reactor design, the critical heat flux (CHF) is one of the most important parameters for the reactor safety analysis. The occurrence of CHF will cause a sharp increase in the fuel rod surface temperature, which will result in the failure of fuel claddings and damage of the core. The CHF depends on the local flow conditions and the geometry of the flow channels, which makes the prediction of CHF in a fuel assembly more difficult when considering the cross flow between neighboring channels, spacer grids and mixing vanes. In this paper, the departure from nucleate boiling (DNB) type CHF in rod bundles under motion conditions is investigated based on the coupled analysis of the subchannel method and a CHF mechanism model, namely the liquid sublayer dryout model. The liquid sublayer dryout model assumes that there is a thin liquid sublayer underneath a vapor blanket formed by the coalescence of small bubbles near the heated wall. The dryout of this sublayer is considered as the CHF occurrence. In the liquid sublayer dryout model, sublayer thickness, velocity and length of the vapor blanket are three crucial parameters. In present research, the subchannel code calculates the local flow conditions for the rod bundle and provides input parameters for the liquid sublayer dryout model to predict CHF. In order to verify the method above, the predicted results are compared with the CHF Look-Up Table 2006 (LUT-2006) and a reasonable agreement can be achieved. In addition, the effects of rod bundle inlet coolant mass flow rate, subcooling and motion conditions on the CHF are analyzed.
The Radiation Protection Design of PWR Spent Fuel Dry Storage Facility
Authors: Huang Liming , Yang Shouhai , Liu Jie
Published: October 2018

Radiation safety is an important part of safety assessment of spent fuel dry storage technology. This paper describes the radiation protection design of PWR spent fuel dry storage facility for radiation safety completed by China General Nuclear Power Corporation. Considering the special site conditions, Monte Carlo method is used to complete the precise calculation of the three-dimensional radiation dose field in the spent fuel storage building. Through the spent fuel storage module and the storage building with shielding function, radiation shielding design is completed to meet China’s regulatory requirements, which ensures radiation safety for workers and the public during the transport and storage of spent fuel. It will provide a reference for construction of spent fuel dry storage facility of CPR1000 and HPR1000.
Comparative Study of Helium Turbine Brayton Cycle and Supercritical CO2 Brayton Cycle for HTGR
Authors: Zhao Gang , Yang Xiaoyong , Ye Ping , Wang Jie , Peng Wei
Published: October 2018

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM), which is designed by Tsinghua university of China, is under construction in Shidao Bay of China. It will be the world’s first pebble-bed type modular HTGR commercial demonstration plant. In HTR-PM project, steam-Rankine cycle has used in the power conversion system because it represents current state-of-the-art technology. Meanwhile, helium turbine for HTGR has been investigated for many years in Tsinghua University. Mock-up machine for HTR-10GT has been built. Helium turbine for 250MW HTGR, which is based on HTR-PM, has completed conceptual designed. However, supercritical carbon dioxide (S-CO2) Brayton cycle has shown to have great potentials for future HTGR technology in recent years because of its critical properties. Helium turbine cycle and S-CO2 Brayton cycle are both candidates for future HTGR. Therefore, comparative study is conducted in this paper. Comparison is focused on achievable efficiencies for each cycle mentioned above and on cycle layout with respect to simplicity and compactness, which primarily determines capital cost. Firstly, the physical model for helium turbine cycle with recuperator and intercooler is built and cycle performance is analyzed based on the parameters of HTR-PM. Then the model for S-CO2 Brayton cycle with recompression is also built and the cycle efficiency is analyzed with the same parameters of HTR-PM. Secondly, comparison between helium turbine cycle and S-CO2 Brayton cycle is made from the view of thermodynamics. Moreover, parameters optimization of both cycles based on HTR-PM is carried out. At last, advantage and drawback of both cycles are discussed from the engineering point. In conclusion, cycle simplicity and technology maturity of helium turbine cycle are better than S-CO2 Brayton cycle. But on the other side, smaller size equipment and less compression work of S-CO2 Brayton Cycle are more competitive than helium turbine cycle. Helium turbine with higher temperature and S-CO2 Brayton Cycle with higher pressure can achieve higher efficiency than steam Rankine cycle.
Residual Unbalanced Mass Determination of an AMBs Controlled Rotor Based on Control Current Analysis of the Feedback Loop
Authors: Fan Tianpeng , Sun Zhe , Zhang Xiaoshen , Yan Xunshi , Zhao Jingjing , Shi Zhengang
Published: October 2018

Active magnetic bearing technology is used more and more for its high performance, such as high speed and frictionless operation. But the rotor vibrates sometimes during operation due to the existence of residual unbalanced mass, which may affect the security of the whole system. In order to determine the distribution of residual unbalanced mass, this paper proposes a method based on frequency response, control current analysis, and image data processing. The theoretical and calculated results show the validity of the method.
Particle Release During Laser Decontamination of Concrete Surfaces
Authors: Kahl Torsten , Greifzu Georg , Herrmann Marion , Lippmann Wolfgang , Hurtado Antonio
Published: October 2018

The phase–out of all Nuclear Power Plants (NPP) until 2022 in Germany offer the opportunity to apply new and efficient decommissioning technologies, which allow further reduction of decommissioning costs and minimization of the collective dose for personnel. One challenging task in this process is the decontamination of protective paints containing polychlorinated biphenyl (PCB). PCB-containing surfaces demand consecutive radiological decontamination and removal of the PCB-containing paints before the demolition is possible. Laser technology, commonly used in many different industrial sectors, presents an advanced approach to this problem.

Successful thermal decomposition of PCB-containing paints using high power diode lasers has been reported by our group in former publications [1]. Ongoing investigations focus on the technology transfer from static laboratory state to industrial application. This includes the verification of safety and efficiency issues for the complete laser supported process.

The current presentation covers all aspects of the laser process: concrete sample preparation as well as an experimental set-up utilizing a mobile laser system that includes an innovative laser head. Particular focus will be on the phenomena of the generated concrete surfaces and the arising by-products during laser processing.

A 10 kW high power diode laser, Laserline LDF 1500-10000 was used for the ablation experiments. Concrete samples were coated with typical decontamination epoxy-based paints, which match the commonly used paint systems in German NPPs in terms of composition, structure and thickness. The influence of significant processing parameters (laser power, feed rate, type of concrete and composition of paint) on the laser ablation are shown in relation to the energy input per length. The release of particles is subjected to these parameters as well. Therefore the size and the shape of the arising particles have been measured online in accordance with these processing parameters by an Engine Exhaust Particle Sizer (EEPS, Model RP-3090, TSI) in a range between 5.6 to 560 nm. The quantity of particles has also been counted by a Condensation Particle Counter (CPC, Model 3022, TSI). Overall up to 6 million particles per cubic decimeter are detected with a dilution factor of 1:100 throughout the experiments. Both measuring systems are connected to the exhaust air pipe downstream, next to the laser head. The shift of the particle size in accordance with the above mentioned parameters is crucial for a comprehensive understanding of the laser decontamination.

The mobile laser supported decontamination technology will be tested in 2018 in a German nuclear facility. This test will contain the decontamination process with actual surface configurations like floors, walls, ceilings and corners, and will act as an effective proof-of-concept for the developed laser system prototype.
Pump Bearing Fault Detection Based on EMD and SVM
Authors: Feng Yi , Li Xianling , Ke Zhiwu , Chen Zhaoxu , Tao Mo
Published: October 2018

In nuclear power plant system, pump is the key equipment to maintain the flow of the primary loop coolant and the secondary loop heat transfer fluid. The main coolant pump and the feed water pump are in long-term operation status. Bearings are the key components to ensure stable operation of the pump, and which could be damaged in abnormal conditions. Once the failure occurred in the bearings, pumps would exhibit periodic vibration, which might cause the flow pulsations of coolant and heat transfer fluid; gradually, these situations could reduce the control accuracy and the stability of pump. Therefore, the detection and diagnosis of pump bearings are significant to improve the safety and stability of reactor system. We proposed an approach combined with signal processing and machine learning to extract the signal features and recognize the signal samples automatically. The proposed approach consists of three main steps: firstly, empirical mode decomposition (EMD) is applied to decompose the signals into several intrinsic mode functions (IMFs) which are corresponding to the different components of the original signals; secondly, calculating the correlation coefficient between each IMF and the original signal, the correlation coefficient sequence imply the components distribution of the signal which can be applied to recognize the signal samples; finally, extracting a part of correlation coefficient sequences to train the support vector machine (SVM), and then an classifier can be obtained and use to recognize the other signal samples automatically. Experimental results show that this method can effectively detect the pump bearing operating conditions and failures, and can provide a reference for the safe and stable operation of reactor pumps.
Example of Graded and Lifecycle Phase-Specific Security Controls for Nuclear I&C and EPS Use Cases
Authors: Watson Venesa , Bajramovic Edita , Lou Xinxin , Waedt Karl
Published: October 2018

Working Group WGA9 of IEC SC45A (Nuclear I&C and ES), has recently completed a further working draft (WD) of the new IEC 63096 (unpublished) standard, aptly entitled Nuclear Power Plants – Instrumentation, Control and Electrical Systems – Security Controls. IEC 63096 specifically focuses on the selection and application of computer security controls for computer-based I&C and ES systems. This standard follows the commonly accepted ISO/IEC 27000 series security objectives of confidentiality, integrity and availability, and borrows and expands the objectives and implementation guidance from ISO/IEC 27002, while considering recommendations on sector-specific standards by ISO/IEC 27009. In addition, this guidance introduces a security grading, as well as lifecycle phase-specific controls. The grading aligns with the stringency of security controls, starting with Baseline Requirements (BR), Security Degree S3 and up to S1 (from lowest to highest degree). The lifecycle phase concerns the I&C development (D), project engineering (E) and operation and maintenance phases (O). This paper applies a sub-clause of IEC 63096 clause 15 (Supplier Relationships), to a programmable logic controller (PLC) that is typically used in power plants, to show the intended use of this standard and how it complements highest safety requirements in power plants. The Supplier Relationship clause concerns topics related to supply chain security, and is used to develop a use case example for the PLC. This example demonstrates how the controls and security degrees fits the implementation guidance from ISO/IEC 27002 and how they can be methodically applied to an I&C system.
Numerical Simulation of Thermal Hydraulic Phenomena During Spray Injection Using Lagrangian and Eulerian Approaches
Authors: Hong TaeHyub , Go JongWook , Seo MiRo
Published: October 2018

In this study, computational fluid dynamics simulation of gas-droplet flow in a spray injection in a THAI vessel is performed and the heat and mass transfer between the gas and the droplets are investigated. The purpose of the numerical simulation was validation of a CFD model to use in a hydrogen risk assessment in a containment building under severe accident conditions. The numerical simulation has to cover the physical phenomena that occur when spray water is injected downward into stationary gas in a large closed vessel. In order to model the two phase flow of the gas and droplets, two different modelling approaches are applied, one based on the Lagrangian approach and the other based on the Eulerian approach. In both approaches, the gas-droplet interactions are modelled using the two-way Lagrangian particle model and dispersed multiphase model to simulate gas-droplet interaction, two way coupling of momentum, and the heat and mass transfer between gas phase and droplet phase. For droplet heat and mass transfer, convective heat transfer and diffusion limited mass transfer are assumed. The calculated change in the temperature of the gas shows qualitative agreement with the experimental results but the gas temperature decrease was over predicted with both approaches. The calculated pressure shows good agreement with experimental results in both approaches. Both approaches also show similar prediction of temperature and pressure; therefore, they can both be applied to the containment analysis. The heat transfer coefficient in the gas-droplet heat transfer should be modified to be suitable for the spray flow. When considering that the spray system in the containment building is to be operated de-pressurized, the CFD model for the spray flow should be suitable for the thermal hydraulic analysis of the containment building.
Numerical Simulations on Hydrodynamic Process of Melt Jet Breakup and Fragmentation With the Two-Phase Lattice Boltzmann Method
Authors: Saito Shimpei , Abe Yutaka , Kaneko Akiko , De Rosis Alessandro , Festuccia Alessio , Koyama Kazuya
Published: October 2018

Jet breakup and fragmentation are important phenomena to be well understood during a core-disruptive accident of sodium-cooled fast reactors. The three-dimensional two-phase lattice Boltzmann model developed previously by the authors is improved in numerical stability used to simulate the hydrodynamic process of melt jet breakup. Nonorthogonal central moments is successfully introduced into the model. Numerical simulations of FARO-TERMOS experiments demonstrate the enhancements in stability of the present model. The simulations with two types of grid resolutions show the effect of spatial resolution on the results.
The Design of a Nuclear Emergency Decision Deduction and Training Platform
Authors: Ke Li , Manchun Liang , Guofeng Su , Jie Yang , Jingtao He , Shuijun He
Published: October 2018

Nuclear accident and nuclear terrorism are of great threats to international security. The countermeasures against nuclear emergency situation should be deliberately considered before it happened. In accordance with experience from Fukushima Daiichi Accident, various kinds of complex situations must be considered at the phase of emergency preparedness. And all related schedules, including national laws and regulations, allocation of responsibility and authority, emergency plans and procedures, facilities, instruments, trainings and exercises must be examined. The appropriateness of the emergency preparedness and the emergency response capability need to be verified through emergency exercises. However, exercises are usually costly and labor-intensive. So, it is necessary to design a software system to help conduct emergency exercises simulation.

A Nuclear Emergency Decision Deduction and Training Platform (NEDDTP) is designed in China Nuclear Security Center of Excellence, which aims at performing the simulation of nuclear emergency process and providing a training platform for emergency workers. 3D models, the environment of nuclear power plant (NPP), some physical numerical models for simulating and several typical nuclear emergency scenarios for training are included in the platform. NEDDTP realizes the function of digitization of scenario, emergency simulation, resource management, command and dispatch, process record and evaluation in exercise, and it also provide VR display for emergency workers.
Calculational-Experimental Monitoring of Radiation Damage Parameters on VVER Equipment and Their Application During Equipment Residual Life-Time Estimation
Authors: Borodkin Pavel , Gazetdinov Azamat , Khrennikov Nikolay
Published: October 2018

The problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.
Research on Algorithm of Sump Level Operator Assisted Support Program for PWR Nuclear Power Plant
Authors: Qiaofen Liu , Sanping Xiao , Yu Liu , Xichao Liu , Xulun Jiang
Published: October 2018

Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.
Experimental Study on the Critical Heat Flux of Nanofluid Flow Boiling Under Different Conditions
Authors: Wang Y. , Deng K. H. , Wu J. M. , Yue N. N. , Zan Y. F. , Su G. H. , Qiu S. Z.
Published: October 2018

Nanofluid has been attracted great attention since it was proposed as a preeminent working fluid. Flow boiling is familiar in heat transfer system and the critical heat flux is a key parameter for the design of thermal hydraulic. In present work, the critical heat flux of nanofluid flow boiling is experimentally investigated in a vertical tube with the consideration of outlet pressure, mass flux, inlet subcooling, heating length and diameter. The results indicate that the critical heat flux of nanofluid flow boiling is enhanced compared with base fluid and the increasing radio is increased with increasing the mass flux, diameter and pressure, and with decreasing the heating length. In addition, the inlet subcooling and concentrations (0.1vol.%, 0.5vol.%) have almost no significant influence. Furthermore, a new mechanism for the enhancement of nanofluid flow boiling critical heat flux was proposed by the SEM images of nanopariticle deposition on the heating surface.