bluuebunny/crossref_metadata_2025 huggingface dataset:
Preliminary Design for the Back Plate of Sub-Module of CN HCCB TBM
Authors: Zhao Zhou , Hu Gang , Wang Qijie , Cao Qixiang , Feng Kaiming , Zhang Guoshu
Published: February 2014

The latest design of CN HCCB TBM is 1×4 configuration, which is consisted of 4 sub-modules and a big back plate. For the sub-module, the back plate is the most key and complicated part that is a multi pipes and plates structure, the space between the plates are to form the functional cavities that are used to distribute and collect the coolant and purge gas, the cavities are connected by the pipes. The cavity 1 is used to distribute the coolant (from the big back plate) into the branch cooling channels of caps and rib, the cavity 2 is to collect the coolant from the branch channels of caps and rib and to distribute it into lots of branch channels of cooling plates, the cavity 3 is to mix and collect the coolant from the branch channels of cooling plates that will flow into the back plate of TBM finally.

The preliminary analysis results show that, the mass flow rate is between 2.20×10−2 and 2.33×10−2 kg/s for the coolant flowing into branch channels of caps, and it is between 1.82×10−3 kg/s and 1.84×10−3kg/s for that of rib, which indicates the cavity 1 will lead a good uniformity of mass flow rate; although the mass flow rate is between in a large range from 5.29×10−4 kg/s to 9.19×10−4kg/s for that of cooling plates, the thermal hydraulic analysis of sub-module shows that the minimum value (5.29 × 10−4 kg/s) can also meet the temperature requirements of materials.
Improvement on Dual Energy CT Reconstruction Method With Reduced Data
Authors: Liu Yuanyuan , Zheng Peng , Zhang Chunming
Published: February 2014

Dual energy CT (DECT) has become a hot topic for its high detection precision and robust material identification ability in the field of nuclear safety and security inspection. However, the high cost of the system becomes a big limitation for its wide usage. To solve this problem, in 2009, we have proposed a dual energy CT reconstruction method with reduced data (DECT-RD) requiring much fewer data to reduce the cost of detectors. However, it is a simple idea without more analyzing in the process of solving ill-posed equations. In this paper, we tried to solve ill-posed equations with constraint condition (DECT-RDCC) and least squares (DECT-RDLS) respectively. Numerical simulations are done by using DECT-RD, DECT-RDCC and DECT-RDLS in the same situation, only 7 dual energy detector bins instead of 256 complete bin sampling in each projection. Results demonstrated that DECT-RDCC with relative error less than 1.1% is better than DECT-RD with relative error less than 1.8% while DECT-RDLS plays a more exact and steady role with relative error less than 0.6% than DECT-RDCC. Hence, DECT-RDLS is a better method used to obtain much lower system cost. We believe this work will drive DECT into wide usage.
Consideration on Selection of Design Codes and Standards for China Fusion Engineering Testing Reactor
Authors: Li Guoqing , Xing Chao , Kang Yexi , Li Xiaozhen
Published: February 2014

After establishment of national integration design group for magnetic confinement fusion reactor in 2011, China has started its concept design activities for China Fusion Engineering Testing Reactor (hereinafter referred to as CFETR). According to the design goals of CFETR, it will be a nuclear facility contain self-sustained tritium cycle loop. As a nuclear facility, in order to assure the safety and reliability of design results of CFETR, all design should be based on existing codes and standards, or some special specifications. This paper will give introductions to existing major codes and standards in the field of magnetic confinement fusion, including the codes published by American Society of Mechanical Engineers (ASME), the standards published by French society for design and construction and in-service inspection rules for nuclear islands (AFCEN), the technical documents issued by International Thermonuclear Experimental Reactor (ITER) Organization, the standards published by nuclear industry and relating industries in China, and so on. After taking into account the requirements of the CFETR and the status of standardization of nuclear industry in China, the paper will analyze and discuss the considerations on how to select applicable design codes and standards for CFETR.
Thermal Hydraulics Numerical Simulation and Optimized Design of Passive Residual Heat Removal Heat Exchanger in AP1000
Authors: Xia Huining , Lu Daogang , Du Zheng , Zhang Yuhao , Fu Xiaoliang , Ma Zhongying , Yang Yanhua , Liang Min
Published: February 2014

A model of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and In Containment Refueling Water Storage Tank (IRWST) in AP1000 had been scaled by power-volume theory. FLUENT code analysis for the characteristic of thermal stratification phenomenon and natural convective of the scaled model was performed for 2000s of transients to investigate the temperature field distributions and flow field distributions in IRWST. As for the numerical experiment, the distributions had been achieved at different times.

The centralized layout of PRHR HX, which looks like a circle in a section, makes the temperature of fluid outside tubes higher at the location that closer to the upper of the circle in the horizontal tube section and closer to the center of the circle in the vertical tube section, and there may appear nuclear boiling. Then temperature difference between the tube side and fluid side is smaller, causing the heat exchanger efficiency is worse at these locations.

In order to improve the heat transfer effect of PRHR HX, two schemes had been proposed. First, increase horizontal section length and shorten the vertical section length. Second, Flat layout replaces centralized layout. These models were established and simulated by FLUENT. Heat transfer coefficients of “C-tubes” were comparative analyzed to study which kind of the heat transfer effect was better. It contributes to provide some references for improving the heat transfer effect of PRHR HX.
Numerical Simulation of Self-Leveling Behavior in Debris Bed by a Hybrid Method
Authors: Guo Liancheng , Morita Koji , Tagami Hirotaka , Tobita Yoshiharu
Published: February 2014

The postulated core disruptive accidents (CDAs) are regarded as particular difficulties in the safety analysis of liquid-metal fast reactors (LMFRs). In the CDAs, the self-leveling behavior of debris bed is a crucial issue to the relocation of molten core and heat-removal capability of the debris bed. The fast reactor safety analysis code, SIMMER-III, which is a 2D, multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model, was successfully applied to a series of CDA assessments. However, strong interactions among rich solid particles as well as particle characteristics in multiphase flows were not taken into consideration for fluid-dynamics models of SIMMER-III. In this article, a developed hybrid method, by coupling the discrete element method (DEM) with the multi-fluid model of SIMMER-III, is applied in the numerical simulation of self-leveling behavior in debris bed. In the coupling algorithm the motions of gas and liquid phases are solved by a time-factorization (time-splitting) method. For particles, contact forces among particles and interactions between particles and fluid phases are considered through DEM. The applicability of the method in such complicate three phase flow is validated by taking the simulation of a simplified self-leveling experiment in literature. Reasonable agreement between simulation results and corresponding experimental data shows that the present method could provide a promising means for the analysis of self-leveling behavior of debris bed in CDAs.
Standardization of 243Am by Efficiency Extrapolation Method
Authors: Zhong Zhen-yuan , Bai Tao , Li Mou , Liu Jie , Dai Yi-hua , Shang Jian-bo , Lei Bei-fang
Published: February 2014

243Am is a very important nuclide as a chemical yield tracer for the radiochemical determination of 241Am in environmental, food and other samples. For the study of absolute activity measurement of 243Am samples, efficiency extrapolation method by 4πα,β(PC)-γ(NaI(TI)) counting system is introduced in this paper. A solution containing 243Am in equilibrium with its daughter 239Np was standardized by this coincidence system, and the activity concentration of 243Am was just half of the gross activity concentration based on the theory of radioactivity equilibrium. In the experiment, the efficiencies of 243Am and 239Np were changed by altering the applied HV step by step, and the gross activity was obtained while the total efficiency of α and β was extrapolated to 100%. The final result for the efficiency extrapolation method is 5116.1(1±0.89%) Bq/g, which is the mean of the results obtained for the eight VYNS sources. As comparison, the result for ionization chamber is 5141.2(1±0.39%)Bq/g which is the mean of the results obtained for the six glass sources. The results show that there is good consistency between both methods adopted, and the agreement between results by both expereiments was 0.49%. The reference time is January 1 2012 0h.
Comparison of Different Acceptance Standard for Radiographic Rounded Indications in Welds
Authors: Wang Cong , Tang GuoXiang , Zhao JiaGuan , Yang HengLiang
Published: February 2014

Usually, we do radiographic testing according to different standards. Evaluation and acceptance of rounded indications in different standards are different from each others. This paper will evaluate some radiographs based on ASME BPVC, RCC-M and GB/T3323 respectively, and then a conclusion will be drawn that ASME BPVC in rounded indications’ evaluation and acceptance commonly rely on subjective judgment of radiographs evaluator, lack of quantitative evaluation.

This paper will make a comparison on rounded indication evaluations and acceptances in ASME BPVC, RCC-M and GB/T3323, exploring their differences.
Tripoli-4 Gamma-Ray Dose Calculation for Spent PWR Fuels
Authors: Lee Yi-Kang , Sharma Kabir
Published: February 2014

The gamma-ray dose calculation is essential for the radiation shielding of pressurized water reactor (PWR) spent fuels. Homogenization modeling of fuel pin lattices for typical PWR spent fuel pins is regularly applied on the radiation protection calculation of gamma-ray dose in an air medium. However, depending on the size of the homogenized lattice and the location of the detectors, under-estimation or over-estimation of the gamma-ray dose due to the homogenization modeling can be obtained with respect to the detailed heterogeneous model. In previous published results from MCNP-4A and 4C calculations on gamma-ray dose from spent PWR fuel pins, very different homogeneous to heterogeneous (Hom/Het) ratios were reported. In this study these Hom/Het ratios have been re-evaluated and benchmarked by using the TRIPOLI-4 Monte Carlo transport code. The new TRIPOLI-4 mesh tally capabilities have also been applied to calculate the radial and axial gamma-ray dose distribution. With the recently upgraded TRIPOLI-4 display tool, the dose rate maps and the isodose rate curves around a spent PWR fuel assembly have been established.
Proton Dose Conversion Coefficients Based on Chinese Reference Adult Woman Voxel Phantom
Authors: Liu Fangfang , Shen Mingqi , Li Taosheng , Liu Chunyu
Published: February 2014

In order to calculate the dose conversion coefficients for proton, the voxel model of Chinese Reference Adult Woman (CRAW) was established by the Monte Carlo transport code FLUKA according to the Chinese reference data and the Asian reference data. Compared with the reference data, the deviations of the mass for organs or tissues of CRAW is less than ±5%. Calculations have been performed for 14 incident monoenergetic protons energies from 0.02GeV to 10TeV at the irradiation incident of anterior-posterior (AP) and posterior-anterior (PA). The results of fluence-to-effective dose conversion coefficients are compared with data from the different models such as an anthropomorphic mathematical model, ICRP reference adult voxel model, the voxel-based visible Chinese human (VCH). Anatomical differences among various computational phantoms and the spatial geometric positions of the organs or tissues lead to the discrepancies of the effective dose conversion coefficients in the ranging from a negligible level to 107% at proton energies below 0.2GeV. The deviations of the coefficients, above 0.2GeV, are mostly within 10%. The results of fluence-to-organ absorbed dose conversion coefficients are compared with the data of VCH. The deviations of the coefficients, below and above 0.2GeV, are within 150% and 20%, respectively. The primary factors of the deviations for the coefficients should be due to the differences of the organ mass and the size of the body shape.
Falling Characteristic Analysis of Passive Drive Mechanism of Absorber Ball Shutdown System for HTR-PM
Authors: Chen Feng , Li Tianjin , Liu Yuan , Huang Zhiyong , Bo Hanliang
Published: February 2014

The absorber ball shutdown system is a key system and related to reactivity control in HTR-PM, which has once achieved power regulation and cold shutdown of the nuclear reactor. Passive drive mechanism of absorber ball shutdown system has been designed for HTR-PM reactor safety before. Ball dropping valve can be automatically opened by gravity when loss of offsite power happens, then a mass of balls with BC4 fall into reactor core for function implementation. Therefore drive mechanism reliability is fairly important for absorber ball shutdown system in HTR-PM. Experiments on passive drive mechanism have been carried out in this study. Falling distance and falling time of drive mechanism have been recorded and analyzed with and without absorber ball, as well as at ambient temperature and in the heating conditions. Experimental results demonstrate that designed passive drive mechanism can shift reliably. Ball dropping valve can freely open each time without absorber balls in the storage vessel. Basically present design of passive drive mechanism for absorber ball shutdown system can meet the requirement of actual reactor application.
Oxidation/Dissolution of Ferritic ODS Steels in Static Lead With Various Oxygen Contents at 650°C
Authors: Tsisar Valentyn , Yeliseyeva Olga , Konys Jürgen
Published: February 2014

The compatibility of ferritic oxide dispersion strengthened (ODS) steels with static liquid Pb at 650°C was investigated for up to 1750 h. The dependence of corrosion behavior on oxygen concentration in the melt and steel structure was determined. The interaction mode changed from intensive oxidation in the oxygen-saturated melt, accompanied by the formation of a bi-layer magnetite/spinel scale, to formation of a thin Cr-based oxide film in oxygen-doped melt and to the severe steel dissolution in oxygen-free lead. The viability of the in-situ formed Cr-based oxide film, as a promising way to improve the corrosion resistance of steels in the high-temperature (T > 550°C) lead melts is discussed.
A New Model Towards Thermal Mixing and Stratification in Passive Containment
Authors: Zhang He , Niu Fenglei , Yu Yu , Chen Peipei
Published: February 2014

Thermal mixing and stratification often appears in passive containment cooling system (PCCS), which is an important part of passive safety system. So, it is important to accurately predict the temperature and density distributions both for design optimization and accident analysis. However, current major reactor system analysis codes only provide lumped parameter models which can only get very approximate results. The traditional 2-D or 3-D CFD methods require very long simulation time, and it’s not easy to get result. This paper adopts a new simulation code, which can be used to calculate heat transfer problems in large enclosures. The new code simulates the ambient fluid and jets with different models. For the ambient fluid, it uses a one-dimensional model, which is based on the thermal stratification and derived from three conservation equations. While for different jets, the new code contains several jet models to fully simulate the different break types in containment. Now, the new code can only simulate rectangular enclosures, not the cylinder enclosure. So it is meaningful for us to modify the code to simulate the actual containment, then it can be applied to solve the heat transfer problem in PCCS accurately.
Thermal Analysis of the Head of the Direct Action Solenoid Valve
Authors: Liu Qian-feng , Bo Han-liang
Published: February 2014

Hydraulic Control Rod Drive Technology (HCRDT) is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University owns HCRDT’s independent intellectual property rights. The Integrated valve which is made up of three direct action solenoid valves is the key part of this technology, so the performance of the solenoid valve directly affects the function of the integrated valve and the HCRDT. On one hand, based on the conditions occurring in the operation of the Control Rod Hydraulic Drive System, the thermal characteristics of the head of the valve are analyzed by ANSYS. On the other hand, the numerical results are verified by the experiment. The results show that the temperature of the head of the solenoid valve increases with the current increasing firstly. Second, the highest temperature of the head is obtained so the current can be selected. Third, the coefficient of thermal conductivity of the head of the valve is obtained, which can be applied to other analysis of the solenoid valve.
Scaling Analysis of Liquid Lead–Bismuth Eutectic Thermal Stratification in ADS Facility Under LOCA
Authors: Su Ziwei , Zhou Tao , Li Yunbo , Wu Yican
Published: February 2014

In paper the thermal stratification of liquid lead–bismuth eutectic in accelerator driven subcritical system (ADS) under LOCA is investigated. By means of establishing the space geometric distribution model of thermal stratification which produced in ADS interspace as well as using H2TS scaling analysis to simulate the phenomenon, the main parameters of model facility can be calculated and discussed. When the geometric scaling ratio is chosen be 1/10, by analyzing the Ri number, we can get scaling ratio of leak fluid velocity U0 and process time τ in model is 1/3.16 of the prototype, leak fluid flow Q0 and power q is 1/316.23 of the prototype in the formation process of thermal stratification. Moreover, scaling distortion quantitative analysis for the scaling simulation results be made. So the law related to the time, speed change and so on can be researched, all these work would provide reference and theory for testing facility constructing and ADS safety analysis.
Application of Coincidence Technique for Neutron Flux Mapping in Boron Concentrated Paraffin Wax
Authors: Mehboob Khurram , Cao Xinrong , Ali Majid
Published: February 2014

In this paper, Neutron Activation Technique (NAT) is used for neutron flux mapping in homogenous boron concentrated paraffin wax. For this purpose, activator detectors were prepared by irradiating process in an Americium-Beryllium (241Am–9Be) neutron field for the required time period to induce approximate saturation activity. After this, the induced activity was measured by time coincidence method using a Sodium iodide (thallium) NaI (Tl) crystal detector and organic scintillation “Anthracene” detector for gamma ray and beta particle detection respectively. The activator detectors were activated at different radii from the irradiator neutron source in boron mixed paraffin wax. The flux profile was mapped by measuring induced activity. Errors in measurements and calculations were estimated by Poisson distribution.
Inspection Qualification of Ultrasonic Inspection for RPV Safe End Narrow Gap Welding With Visualization of Laser Ultrasonic Detection Technology and Computerized Ultrasonic Simulation
Authors: Yuan ShuXian , Li Ming , Ma GuanBing , Xiao XueZhu , Lv TianMing
Published: February 2014

According to the method used in ENIQ (European Network of Inspection and Qualification), the performance demonstration of automatic ultrasonic inspection on the narrow-gap welds between the safe-ends (forging) and primary pipe (casting) in one PWR Nuclear Power Plant in China, was successfully finished. The visualization of the laser ultrasonic detection technology — which was introduced for the first time — and ultrasonic simulation technology, were used to study the dynamic propagation of the ultrasonic beam in the narrow-gap welds, in order to provide technical support for performance demonstration and field inspection.
APR+ Core Flow and Pressure Distributions Under the 4 Pump Unbalanced Flow Condition
Authors: Kim K. , Euh D. J. , Youn Y. J. , Chu I. C. , Choi H. S. , Kwon T. S.
Published: February 2014

The core inlet flow rates and exit pressure distributions of an APR+ (Advanced Power Reactor Plus) reactor were evaluated experimentally in this study. The tests were performed in the ACOP (APR+ Core Flow & Pressure) test facility constructed with a linear reduced scale of 1/5 referring to the prototype plant. The major flow path inside the reactor vessel was designed with a preservation of a geometrically similar flow without hindering the dynamic similarity. The 257 core simulators with 771 pressure impulse lines were installed in the ACOP facility to measure the hydraulic characteristics at the inlet and outlet of the fuel assemblies. The pressure distributions along the major flow path were obtained by measuring the static pressure and differential pressures at 584 points. The hydraulic characteristics of the reactor flow under an unbalanced cold leg flow condition were investigated by using an ensemble averaging process of 5 independent tests. The details of these experiments and a data analysis were described in this paper.
Influence of the Fragmentation Behavior on Molten Material Jet Breakup in Coolant
Authors: Iwasawa Yuzuru , Abe Yutaka , Kaneko Akiko , Kuroda Taihei , Matsuo Eiji , Ebihara Kenichi , Sakaba Hiroshi , Koyama Kazuya , Itoh Kazuhiro , Nariai Hideki
Published: February 2014

For the safety design of a Fast Breeder Reactor (FBR), Post Accident Heat Removal (PAHR) is required when a hypothetical Core Disruptive Accident (CDA) occurs. In PAHR, it is strongly required that the molten core material can be solidified and cooled down by the sodium coolant in a reactor vessel. In order to estimate whether the molten material jet is completely solidified by sodium coolant, it is necessary to understand the interaction between the molten core material and the coolant. The objective of the present study is to clarify the dominant factor of the jet breakup length and the size of the fragments experimentally. In this study, we injected molten material (Sn–Bi alloy) into coolant (water) at free fall speed. We can simulate an actual FBR system by using Sn-Bi alloy and water because the density ratio of them is similar to that of an actual FBR system. The jet breakup and the fragmentation behavior of the molten material jet were observed with a high speed video camera. In the previous study which we conducted, solidified crust is generated by the solidification on the molten material jet surface and affects the jet breakup and the fragmentation behavior. Then from the experimental results, in order to predict the size of fragments, it is constructed that the instability model based on hydrodynamic and material mechanics. Then in this paper, the surface close-up of the molten material jet was observed in order to investigate the effect of the solidification on the molten material jet surface.
Study on Helium Leak Detecting for Post Irradiation Fuel Rods
Authors: Kuang Liuwei , Jiang Linzhi , Ren Liang , Yu Feiyang , Guo Chengming , Yin Chunyan
Published: February 2014

In this paper, many methods on testing for post irradiation fuel rods were introduced, especially the method which used helium mass spectrometer leak detector. Helium detection device including pressure helium device, sniffer probe of helium detection device and vacuum chase leaks of helium detection device were designed. First, the device which tested and verified during the mockup experiment had been verified effectively. Then, a series of helium detection experiments were carried out for post irradiation fuel rods in hot cell. The experiments which was divided into three stages; before cutting, leak detection was carried out with sniffer probe of helium and vacuum chase leaks of helium, the background approach to leak rate, which indicated that there was no leakage or helium had been completely escape. When cutting, greater pressure change caused by helium which released from fuel rods, The result showed that there was no leakage. After cutting, in the following helium detection experiment, the background still approach to leak rate, which also indicated that there was no leakage. All the experiments indicated that helium leakage device were effective in great extent, in short, helium leakage device and method were good useful for leakage detection in hot cell, helium leak detecting for post irradiation fuel rods became true.
The Computer Measuring and Controlling System for Temperature Field of the Thermal-Shock Test at R1/R2 Test Section in ESPRESSO Loop
Authors: Zhao Fei , Wu Chunliang , Shen Fengyang , Gu Chunxing
Published: February 2014

The temperature field at R1/R2 test section in ESPRESSO loop was usually built by the hand-tune way in the past. In this paper, a portable computer measuring and controlling system for temperature field was established to changing the traditional way. And the curves of temperature field, in the situation that the temperature difference were 80°C and 170°C simulate CEFR in-pile in sodium environment of R1/R2 test section, had been set up successfully through the system. Then a thermal shock test was carried out on this basis, and the test purposes were achieved through experiments, which provided necessary condition for subsequent measurement of shape and displacement in core components. It was proved that the system was under good performance by debugging and trial run.
The Computer Measuring and Controlling System for the Plugging Meter of the Mobile Sodium Purification Device
Authors: Zhao Fei , Shen Fengyang , Wu Chunliang , Liu Yunyan , Sun Yu
Published: February 2014

The plugging meter of mobile sodium purification device is the special equipment for on-line monitoring the plugging temperature, and according to which to calculate oxygen content in the sodium. This paper introduced the working principle of the test sodium loop, plugging meter and the computer measuring and controlling system. The core of the system was a data acquisition card, and using pure PID control algorithm to complete the whole measuring process of the plugging temperature and solving the plugging temperature by curve fitting method. The system could also display in real time the change curve of flow and temperature in plugging area, and record the test data simultaneously. After multiple debugging and trial running, the system was at good performance which laid foundation on debugging and running for other plugging meters.
Research on Corrosion and Precipitation Behaviors in LBE Systems
Authors: Tian Li , Nui Fenglei , Gao Sheng , Wu Yican
Published: February 2014

Liquid metal, mainly lead and lead-bismuth eutectic (LBE) has been widely applied in the advanced nuclear reactors and accelerator-driven transmutation (ADS) systems for a series of excellent physical and chemical properties, but LBE is very corrosive to the structural materials such as iron and stainless steel in medium and high temperature circumstance which limits its development seriously. Research works of anticorrosion for LBE has been carried out all over the world. In this paper, A kinetic model for corrosion and corrosion products precipitation in non-isothermal LBE circuits is developed to predict corrosion and precipitation behaviors, the mass transfer convection-diffusion equation is established and the corrosion/corrosion products precipitation rate along the circuit is predicted considering the effects of temperature profile of the entire circuit which are different from local parameters, the overall estimate of the corrosion/precipitation situation is obtained, at the end, the solid-phase oxygen control technique is discussed, a program developed for dissolution rate of the PbO particles in the mass exchanger is proposed.
The AP1000® PWR Project Moving Toward Completion in China
Authors: Harrop Grenville
Published: February 2014

The AP1000® pressurized water reactor (PWR) is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission. The initial deployment of this technology has been the construction of dual AP1000 units in each of two coastal sites in the People’s Republic of China (PRC), at Sanmen (Zhejiang Province) and Haiyang (Shandong Province). The contracts for these units were framed to support the PRC’s intention to achieve self reliance in its nuclear supply infrastructure. Westinghouse is implementing its innovative supply chain strategy, “We Buy Where We Build”™, to promote the technology transfer and increasing levels of localization needed as each unit is constructed.

Since the initial contract award in 2007, the Westinghouse Consortium and the purchasers, State Nuclear Power Technology Corporation of China (SNPTC), the Shandong Nuclear Power Company (SDNPC), and the Sanmen Nuclear Power Company (SMNPC) have worked in harmony to build the units using advanced modular construction techniques that reduce construction timescales and associated risks. First-of-a-kind (FOAK) plant components have been manufactured and delivered, including reactor vessels, steam generators, and other safety equipment. With construction and equipment installation in the final stages, the planning and implementation of the pre-operational testing, system turnover, and commissioning are now underway to prepare for fuel load and future commercial operation.
Magnetic Field Analysis of Active Magnetic Bearings for Helium Blower of HTR-PM
Authors: Guo Leilei , Yang Guojun , Shi Zhengang , Liu Xingnan , Yu Suyuan
Published: February 2014

High-temperature gas-cooled reactor-pebble bed module (HTR-PM) of China, based on the technology and experience of the 10MW high-temperature gas-cooled reactor (HTR-10), is currently in the design and experiment phase. The helium blower is the key equipment in the first loop of the HTR-PM.

Active magnetic bearings (AMBs) are replacing ordinary bearings as the perfect sustaining assembly for the helium blower because they have several advantages: they are free of contact, do not require lubrication, are not subject to the contamination of wear, have endurance, and control performance very well. So the AMB is the appropriate supporting assembly.

The rotor’s length of the helium blower is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The maximal radial load of the upper AMB or the lower AMB is about 1950kg, and the maximal axial load is about 4500kg. So the axial and radial AMBs must have enough carrying capacity to support the rotor. It is difficult to design the AMB in limited space. Many factors may restrict the AMB design, for example, clearance between AMB and rotor, electric current, and environment temperature, etc.

In this paper, design of magnetic field will be analyzed. Heat loss of AMB will be discussed. The designing scheme of magnetic field will offer the important theoretical base for AMB design of the helium blower of HTR-PM.
Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code
Authors: Fouet Fabrice , Probst Pierre
Published: February 2014

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters.

A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology.

The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA.

The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.
Preliminary Neutronics Design and Analysis of the Bit Helium Cooling Ceramics Blanket for CFETR
Authors: Li Jia , Liu Songlin , Ma Xuebing , Pu Yong , Chen Xiangcun
Published: February 2014

CFETR is a Tokamak fusion engineering test reactor whose concept design is being developed in China. It is a key issue for breeding blanket design to attain tritium self-sufficiency as one of important missions of CFETR. This paper presents a preliminary neutronics design and analysis employing a BIT (breeder inside tube) type helium cooling ceramics blanket (HCCB) design concept as one of CFETR blanket design candidates. Firstly, 1D reactor model was designed using ceramic breeder Li4SiO4 and beryllium in pebble for multiplier. The primary blanket parameters were optimized to yield the higher tritium breeder ratio (TBR), including the thickness of outboard breeder blanket, enrichment of Li-6 and ratio of Li4SiO4 to Be. Secondly, based on the optimized blanket parameters and plasma parameters, a detailed 3D neutronics calculation model of 22.5° reactor sector was developed, including blanket modules, shield, divertor, vacuum vessel and TF coil. The gap between blanket modules had been taken into account. Finally, a set of nuclear analyses were carried out addressing the key neutronics issues by Monte Carlo neutron-photon transport code MCNP version 5 and the FENDL-2.1 data library.

The preliminary analysis results showed that the global TBR could achieve 1.21 which satisfied the tritium self-sufficiency demand. Nuclear heat, neutronic flux, and distribution of neutron wall loading (NWL) were also analyzed as source terms of the blanket thermal-hydraulics design and reactor nuclear response.
Nuclear Power Plant Passive Residual Heat Removal System Design and Simulation
Authors: Yuan Xiao , Peng Minjun , Xia Genglei
Published: February 2014

The passive safety systems employed in the design of pressurized water reactor (PWR) can accomplish the inherent safety functions and mitigate the consequences of the postulated accidents. In this paper, a passive residual heat removal system (PRHRs) is designed for a certain nuclear power plant. The RELAP5/MOD3.4 code was used to analyze the operation characteristics of the PRHRs. It shows the PRHRs could remove the decay heat from the primary loop effectively, and the single-phase and two-phase natural circulations could respectively establish in the primary circuit and the PRHRs circuit.
Experimental Investigation on Bubbly Flow in Rectangular Channel Under Rolling Condition
Authors: Yan Chaoxing , Yan Changqi , Sun Licheng , Wang Yang , Tian Daogui , Jin Guangyuan
Published: February 2014

Experimental study of rolling effects on characteristics of bubbly flow in a rectangular duct (43 mm×3.25 mm×2000 mm) was performed under ambient temperature and pressure. The characteristics of bubbly flow in vertical, inclined and rolling channels have been obtained through visualization and photographic recording. In the experiment, the gas and liquid superficial velocities ranged from 0.07 m/s to 0.15 m/s and from 1.6 m/s to 2.39 m/s respectively. The bubble clusters contained no less than two bubbles exist in vertical, inclined and rolling channels. With the inclined and rolling angle increase, the bubble clusters tend to contain more bubbles. In vertical case, most bubbles aggregate in the intervals near the narrow wall, thus forms high number fraction peaks at the position around 2x/w = ±0.5. As the inclination and rolling angle increase, the bubble number fraction near the lower wall decreases and that near the upper wall increases. In addition, the number fraction peak close to the lower wall disappears and that near the upper wall becomes higher and sharper at the inclined and rolling angle no less than 10°. There exist no significant differences in bubble lateral distribution under inclined and rolling conditions at the same angle. To identify the influence of rolling on bubble lateral distribution, comparison between lateral buoyancy and additional buoyancy is made theoretically. The results show that the influence of additional buoyancy on lateral bubble distribution can be neglected result from the magnitude of lateral buoyancy is much larger than that of additional buoyancy.
High Breeding Core of a Supercritical-Pressure Light Water Cooled Fast Reactor
Authors: Yoshida Taishi , Oka Yoshiaki
Published: February 2014

Breeding of plutonium with light water cooling has been studied for many years, but high breeding to meet growing demand for electricity in a developed country has not been accomplished. The purpose of this study is to investigate a high breeding core of Super FBR (supercritical pressure light water cooled fast breeder reactor) with new fuel assemblies consisting of tightly packed fuel rods without gaps, which leads to low coolant to fuel volume fraction. The plant system of a Super FBR is once-through coolant cycle with high head pumps. The coolant flow rate is low due to the high enthalpy rise in the core. It is compatible with the high pressure drop of the new fuel assemblies. Both neutronic and thermal hydraulic design of the core is considered. The challenge of high breeding with light water cooling is to satisfy negative coolant void reactivity, high breeding and low enrichment simultaneously. The core with new assemblies has been designed with the average coolant density of 248 kg/m3. It is achieved by setting 380C inlet and 500C outlet temperature. For satisfying negative void reactivity, a solid moderator layer composed of zirconium hydride (ZrH) rods are adopted in some blanket assemblies. Cross sections of the blanket fuel assemblies with ZrH rods are prepared with assembly-wise calculation, because the pin-wise collision probability calculation overestimates the breeding. MOX fuel is used for seed fuel assemblies.

Three types of core layouts with “radially heterogeneous”, “radiating” and “scattered” seed assemblies have been considered, and “radiating” layout shows best breeding characteristics among them. The seed assemblies in a “radiating” layout are not radially separated so that more numbers of blanket assemblies can be placed in high neutron flux region of a core. Fraction of blanket fuel assemblies with ZrH rods is selected for high breeding.

Super FBR using the new fuel assemblies achieved both negative void and high plutonium breeding.
Improvement of the Flue Gas Cleaning System in Radioactive Waste Incineration Facilities
Authors: Zheng Bowen , Li Xiaohai , Chu Haoran , Jia Chengming , Xu Wei , Yang Lili , Wang Peiyi , Wang Xujin , Yang Liguo , Zhang Xiaobin , Li Chuanlian
Published: February 2014

Three radioactive waste incineration facilities have been built in China. The first one has been in operation for more than ten years. Some problems, such as corrosion and waste water treatment in the flue gas cleaning system, have been found in the long time operation of the first facility. According to the origin of the problems, technical process and equipment material of the flue gas cleaning system have been improved in design and been adopted by the second facilities in China. The results showed that most of the problems were resolved, and the improvement is successful.
Effects of Irradiation Growth on the Thermo-Mechanical Behaviors Evolution in Inert Matrix Fuel Elements
Authors: Gong Xin , Zhao Yunmei , Ding Shurong , Huo Yongzhong , Zhang Lin , Li Yuanming
Published: February 2014

Inert matrix fuel elements consisting of metal cladding and fuel meat with the fuel particles embedded into the inert matrix have higher lifetime compared to the traditional fuel elements used in the present nuclear plants. They have promising prospects to be used in the advanced nuclear reactors and disposal of nuclear wastes, and nowadays are widely used in the research and test reactors. In order to precisely predict their in-pile thermo-mechanical coupling behaviors, the relative simulation methods and several ABAQUS subroutines are built with the following irradiation damage effects involved as irradiation hardening in the metal matrix and cladding, irradiation swelling in the fuel particles and the irradiation growth effect for the cladding. Specially, the irradiation growth of the cladding is modeled in the way of anisotropic thermal expansion, and then its effects on the micro-thermo-mechanical behaviors in the inert matrix fuel elements can be investigated. The comparison of the results for the two cases with and without the irradiation growth shows: (1) irradiation growth will remarkably enhance the mechanical interactions in the cladding, such as the Mises stresses; (2) it will weaken the interaction between the fuel meat and the cladding; (3) the increment of the plate thickness will become a little smaller at low burnup, and then become larger at high burnup.
Numerical Simulation Analysis on the Flow Mechanism of Passive Containment Cooling System
Authors: Song Yang , Li Wei-Hua , Li Sheng-Qiang
Published: February 2014

In the third generation PWR, passive containment cooling system was one of the key security parts. If accident happened, the blasting valves in the containment would automatically open and the steam would be injected into the containment. When the pressure reached a certain value, the tank on the top of containment would automatically spray. Driven by gravity, the cooling water would flow along outer surface of containment. Steam inside the containment would condense on inner wall and the pressure in the containment would decrease.

A numerical simulation of steam jet in the passive containment cooling system was carried out using computational fluid dynamics methods. Flow mechanism in the process was studied. A simplified three-dimensional model of the passive containment cooling system was built. The hexahedral structured mesh was made. The RNG k-epsilon turbulence model was used in solving the transient Reynolds Navier-Stokes equations by the Fluent. The distributions of velocity in containment at different time were acquired. Reference would be provided for understanding of flow mechanism of gas jet-flow in containment during the early stage of accident.
An Investigation of Deterministic Realistic Method for LBLOCA Mass and Energy Release Evaluation for CPR1000 NPPs
Authors: Lin Jianshu , Zhuang Chengjun , Leng Jinzhen , Wang Ting , Zhang Zhenhua
Published: February 2014

With the development of nuclear technology, safety analysis methodologies were advancing from traditional conservative to more realistic approaches. For CPR1000-type NPP, the Conventional Method (CM) which employed MEDUSE, PERSEE and REFLET codes in order was still in use to evaluate the LB LOCA mass and energy release (MER). This complex operation yielded excess conservative result as well as inconvenience. Meanwhile, a deterministic realistic method (MDR) had been applied using Realistic code CATHARE GB for LB LOCA MER analysis in EPR design. In this paper, CPR1000 NPP LB LOCA MER calculation was carried out using MDR method with CATHARE GB code. Calculation results were compared with those obtained through the conventional method to assess the containment pressure margin gain. Further effort was needed to evaluate the feasibility of implementing the new method in CPR1000-type NPP applications.
Risk Assessment of the Intake of Foods and Soil With the Radionuclides and the Air Radiation Dose After the Fukushima Nuclear Disaster
Authors: Fujinaga Aiichiro , Yoneda Minoru , Ikegami Maiko
Published: February 2014

Risk assessment of soil contaminated with radionuclides was performed by considering the intake of radionuclides in foods based on measured concentrations. Due to the disaster at the Fukushima nuclear power plant in 2011, radionuclides were spread out over an area of 13,000 km2. Radionuclides were found in the food, and the Japanese people are concerned about eating foods from Fukushima and the surrounding area. Radionuclides such as 134Cs and 137Cs were found on soil, buildings, plants, and so on. The exposure routes were determined to be (1) food intake, (2) ingestion and inhalation of soil particles, and (3) external radiation from the ground. Then, the total doses of all exposure routes for one year and over a lifetime were calculated, and the committed effective doses for the lifetime were evaluated. To estimate the intake of food, the concentrations can be obtained from the database of the Ministry of Health and Welfare. The foods are divided into 16 types to estimate the daily intake. The geometrical means of the intake were calculated using monthly data for each group of food. Then, the intake from food for one year was calculated for each generation. The committed effective doses were calculated using the intake of each radionuclides times the dose coefficient. For the air dose, three cases were set as exposure scenarios. Case 1 was used for people who stay in a house for 24 hours, such as infants, pregnant females, and bedridden people. Case 2 was used for house wives and office workers, who stay outside for 4 hours and inside for 20 hours. Case 3 was used for children, farmers, and construction workers, who stay outside for 8 hours and inside for 16 hours. As a result, exposure through the ingestion and inhalation of soil particles were negligible, and exposures by food intake and external exposure from the ground were comparatively large. This study shows that the air dose by this disaster should be less than 0.2 μSv/hour to control the radiation dose with the consumption of food being less than 1 mSv/year. However, to maintain the lifetime dose under 100 mSv, several mSv/y is sufficient, considering radioactive decay and dilution by advection and diffusion. The risk assessment based on land use can provide information about the priority of countermeasures against the contamination and provides reasonable decontamination methods or risk management strategies.