bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 10 — Numerical Prediction of Velocity Profile Around Bubble Under Accelerating Condition
Authors: Yoshida Hiroyuki , Nagatake Taku , Takase Kazuyuki , Kaneko Akiko , Monji Hideaki , Abe Yutaka
Published: February 2014

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and shear force through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.

Therefore, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to develop the predictive technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT was expanded to two-phase flow simulation under earthquake conditions. In this paper, the bubbly flow in a horizontal pipe excited by oscillation acceleration and under the fluctuation of the liquid flow was simulated by using the expanded TPFIT. Predicted time series of velocity profiles around the bubbles and shapes of bubbles were compared with measured results under flow rate fluctuation and structural vibration. Predicted results were almost same as measured results qualitatively. And it was concluded that the expanded TPFIT can be applied to qualitative analysis of bubbly flow under accelerating conditions.
Effect of Mo and Bi on Mechanical Properties of Zr-Fe-Cr Alloy at Room Temperature
Authors: Luan B. F. , Yang L. Q. , Wei T. G. , Murty K. L. , Long C. S. , Liu Q.
Published: February 2014

To investigate the effects of Mo and Bi on mechanical properties of a Zr-Fe-Cr alloy at room temperature, seven Zr-Fe-Cr-Mo-Bi alloys with different compositions were designed. They were subjected to a series of rolling processes and heat treatments, and then sampled to measure mechanical properties by hardness and tensile test and to characterize microstructures by scanning electron microscope (SEM) and electron channel contrast (ECC) technique. Results indicated that among them two types of Zr-Fe-Cr-Mo-Bi alloys achieve the designed goals on mechanical properties and have the following advantages: (i) the hardness of the alloys, up to 334HV after annealing, is 40% higher than traditional Zr-4. (ii) The yield strength (YS) and ultimate tensile strength (UTS) of the alloys are 526 MP a and 889 MP a after hot rolling and annealing, markedly higher than the traditional Zr alloy. (iii) Good plasticity of the new Zr-Fe-Cr-Mo-Bi alloy is obtained with about 40% elongation, which is greatly higher than the Zr-Fe-Cr-Mo alloy thanks to the addition of Bi offsetting the disadvantage of addition Mo. Furthermore, according to observations of the microstructure observation, the reasons of the effect of the Mo and Bi elements on the mechanical performance of Zr-Fe-Cr alloy were studied and discussed.
The Development and Validation of Nuclear Cross Section Processing Code for Reactor-RXSP
Authors: Yu Jiankai , Li Songyang , Wang Kan , Wang Guanbo , Yu Ganglin
Published: February 2014

The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor physics calculations. The RXSP(Reactor Cross Section Processing Code) which is developed by REAL (Reactor Engineering Analysis Laboratory) of Department of Engineering Physics in Tsinghua University, has changed the situation in China that nuclear cross section processing has been dependent of NJOY for a long time. The key methods such as fast Doppler broadening, thermal libraries interpolation, and OpenMP parallel acceleration, can be achieved with RXSP. This code is able to process the original data of ENDF/B (Evaluated Nuclear Data File/B) efficiently and accurately to produce the continuous energy point cross section data which is necessary for RMC. By comparing with NJOY, The microscopic and macroscopic verification shows that RXSP has the same accuracy as NJOY while RXSP has saved greatly the processing time to meet the efficient demand in the frequent reactor physics-thermal-hydraulic coupling calculations to solve the complex questions related on a large number of materials and temperature. In addition, RXSP make it available to process the resonance parameters of the R-matrix Limited format.
Research on Chinese Public Mental in Post-Fukushima Era
Authors: Min Li
Published: February 2014

Fukushima nuclear accident aroused large-scale public panic toward nuclear power development. Due to the limit knowledge of nuclear power, some people feel frightened and fearful for nuclear safety, nuclear radiation as well as nuclear accident. As the energy with clean, stable and high-efficiency, nuclear power always takes imperative and irreplaceable role in Chinese energy program. Therefore, public acceptance and basic knowledge towards nuclear power in post-Fukushima era is facing new unprecedented challenge. How to relieve the panic and frightening of the public and recover the confidence of nuclear power safety is gradually becoming the hot issue among public.

This paper makes detailed investigations of current public mental in China toward nuclear power in post-Fukushima era, analyzes the internal and external causes of the panic feeling and further proposes several countermeasures and suggestions accordingly for safe and health development of nuclear power in China.
Young’s Modulus and Gas Tightness Measurement of Ceramic Matrix Composite-SiC for Advanced Reactor Application
Authors: Zhao Yungan , Niu Fenglei , Yu Yu , Zhuo Weiqian , Hao Zulong , Xiao Wei
Published: February 2014

Considering the core outlet temperature of high temperature gas-cooled reactor (HTGR) is above 1000 °C and in the accident conditions, a new type of heat-resistant silicon carbide (SiC) ceramic matrix composites needs to be developed. Analysis of the Young’s modulus and gas tightness of ceramic matrix composite-SiC should be paid attention to due to the inherent brittleness and large porosity of SiC ceramics and also the high temperature and high pressure environment where the fuel cladding works. In this paper, a set of measuring device is designed, which is capable of measuring both Young’s modulus and gas tightness of sic ceramic materials. In order to avoid the shatter of ceramic matrix composite-SiC due to excessive deformation in the measurement process of Young’s modulus, Wheatstone bridge principle is used in the experiment to convert the changes of resistance which generated by the micro-deformation of resistance strain gage adhesive on the surface of the specimen into voltage signal, then the relationship between the amplified voltage signal and the strain of specimen can be established, thereby the Young’s modulus of the specimen can be calculated. The measuring method used in this experiment is more rapid, accurate and economic than previous methods. Besides, there is no need for extra measuring device to measure gas tightness of specimen for it has been taken into account in designing process.
Heat Sink Issue: A Specific Safety Issue for NPP Site Selection and Evaluation in the Warm Countries With Extreme Climatic Conditions
Authors: Rodionov Andrei , Mattei Jean-Marie
Published: February 2014

One of the major issues during the site selection and design of a NPP is the heat sink issue, i.e. installation and availability of reliable and efficient recipient of heat (air, water or both of them) in relation in particular with:

• waste the heat removed from the condenser of the turbines during normal operational state,

• waste the heat removed from all of the safety-related and non-safety-related equipment and systems during normal operational state,

• waste the residual heat removed from nuclear reactor fuel (situated in reactor vessel or tubes), the decay heat removed from nuclear spent fuel (situated in spent fuel pool), and the heat removed from all of the safety-related equipment and systems during abnormal operational state and accident conditions.

There are several “newcomer” countries that are situated in warm climate zones. In such countries, the cooling water design could be a challenge with regards to extreme water and air temperatures, air humidity, absence of an important reliable cold water source inside the country, etc. In addition, some specific items related to natural, industrial and human environment should be considered during the site evaluation, as for example, risk of intensive biota development, oil spills, siltation, low water level events, etc.

This requires particular consideration of the heat sink issue for the site regulatory review during the licensing process. The article discusses the methodology for site regulatory review with regards to the heat sink issue and provides some practical advises on safety assessment and acceptance criteria to be considered.
Spent Fuel Pool Accidents in a Nordic BWR
Authors: Könönen Niina E.
Published: February 2014

Loss-of-pool-cooling accidents at the spent fuel pools in the reactor hall of a Nordic BWR have been studied using the MELCOR 1.8.6 code. Studies were made with several different calculation nodalizations. Other investigated variables were the total decay heat power of fuel assemblies in the pool, the initiator of the accident (loss of pool cooling or loss of coolant from the pool), the LOCA leak elevation, the alignment of the re-flooding injection and the use of a lid on top of the pool.

From the results it was observed that ensuring natural circulation of air in the fuel is essential in preventing fuel damages. In cases where the air flow is prevented (loss of pool cooling, LOCA break elevation above the bottom of fuel assemblies or a lid on top of the pool) the fuel is damaged in nearly all cases. Only with the pool decay power being sufficiently low (2227 kW) the melting of the fuel was prevented.

If water injection to the SFP can be restarted before the fuel temperature reaches the cladding failure criterion (1173 K), the rewetting of fuel assemblies is successful and the fuel temperatures will quickly decrease and the assemblies will remain intact. The re-flooding of already damaged fuel assemblies will also cool the fuel rods and reduce the radioactive releases to the environment. However, it may result in greater hydrogen releases.
Exploration of Bilingual Teaching Mode for Nuclear Specialty
Authors: Chen Min , Li Leimin , Lu Chunhai
Published: February 2014

Bilingual education is the important method of training capability of international cooperation, communication and competition. Compared with other fields, the nuclear specialty needs more professionals. Four kinds of successful methods in foreign countries: transitional bilingual education, maintenance bilingual education, an immersion bilingual education and dual-rail bilingual education are studied and compared in this paper. The difference in bilingual education mode at home and abroad is analyzed. Current situation of bilingual education in China is presented. Drawing lessons from another country, we suggest a proper bilingual education method for the nuclear specialty. We think that “Step by step” approach is consistent with China’s national conditions.
The Effect of Electromagnetic Field on the Behavior of Bubbles
Authors: Liu Hong-bo , Pan Liang-ming
Published: February 2014

The coalescence and motion behaviors of bubbles can be influenced by electromagnetic field, it can be found in the TOKAMK fusion reactors of ITER program. In this paper, VOF (Volume of Fluid) method is adopt to simulate the single bubble rising behavior and bubble coalescence in stagnant liquid under the electromagnetic field within a cylinder closure (D = 20mm, H = 50mm). A three-dimensional numerical simulation is presented considering the driving effect of uniform electric field and magnetic field on the process of bubble rising and bubbles inline coalescence in stagnant liquid pool. The interesting local Magneto-hydrodynamic (MHD) flow, which is produced due to current non-homogenous distribution around the insolated gas bubble, is discussed in the process of single bubble rising and coalescence of bubbles inline. Under the influence of local MHD flow, the rising bubble shape presents obvious deformation. More importantly, as the existence of rotating Lorentz force, the liquid film between two bubbles is stirred, the process of coalescence is largely accelerated.
Experimental Evaluation of Debris Bed Characteristics in Particulate Debris Sedimentation Behaviour
Authors: Shamsuzzaman M. , Horie Tatsuro , Fuke Fusata , Kai Takayuki , Zhang Bin , Matsumoto Tatsuya , Morita Koji , Tagami Hirotaka , Suzuki Tohru , Tobita Yoshiharu
Published: February 2014

After a core-disruptive accident (CDA) in sodium-cooled fast reactor, degraded core material can form debris beds on core-support structure or in the lower inlet plenum of the reactor vessel. This paper reports an experimental evaluation on debris bed formation characteristic in CDA. Investigation of debris bed characteristic during debris sedimentation on core catcher plate is important from recriticality concern and also from cooling considerations to ensure the safety of the reactor main vessel in CDA. In the present study, to evaluate this characteristic, a series of experiments was performed by gravity driven discharge of solid particles as simulant debris from a nozzle into a quiescent water pool in isothermal condition at room temperature. The discharged solid particles with a maximum amount of 10 L finally accumulate on the debris tray, forming a bed with a convex or concave mound depending on the experimental parameters. The nozzle diameter, nozzle height, debris density, debris diameter and debris volume are taken as the experimental parameters. Currently, three types of spherical particles, namely Al2O3, ZrO2 and stainless steel (SS) with diameter of 2, 4, or 6 mm are employed to study the effect of key experimental parameter on debris bed mound shape. In addition, 2 mm non-spherical particles of SS were also utilized to investigate the effect of debris shape on altering mound profile. In experimental investigation with different debris volume, both developing and fully developed mound shapes were observed based on the effect of debris size, density and nozzle diameter. In this study, the investigated particle velocity of main stream settling particles was found increasing with nozzle diameter, which caused a decrement of mound height with an increment of mound dimple area. In nozzle height effect, shrinking of concavity on mound shape was observed with decreasing manner of impact velocity while height is reducing. From the visualization results of the experimental investigations, transformation of bed shape from convex to concave was observed with increasing repose angle incase of 4 mm Al2O3 particle. In general, transformation of bed shape was observed by increasing either nozzle diameter or particle density for all particle type. The present results could be useful to validate numerical models and simulation codes of particulate debris sedimentation.
Development of a Two-Dimensional Modularity Characteristics Code for Neutron Transport Calculation
Authors: Liang Liang , Wu Hongchun , Cao Liangzhi , Zheng Youqi
Published: February 2014

The method of characteristics (MOC) has been widely used in lattice code for its high precision and easy complement. However, the long characteristics method needs large quantity of PC memory when dealing with large scale problems. The modularity MOC method could significantly reduce the PC memory when calculating the problem which contains lots of repeatedly geometries, like the fuel assembly in the reactor. In this method, only typical geometric cells are selected to trace the rays, and then the geometry information of these cells is stored. So, the modularity MOC method is feasible to perform well in the calculation with large scale. When tracing the rays, the technique of mesh ray generating and the corresponding azimuthal quadrature set are both applied. The techniques make sure that each ray has the reflected ray in the boundary so it is convenient to describe the boundary condition. The optimal polar angle and the Guass quadrature set are selected as the polar quadrature set. Furthermore, the coarse mesh finite difference (CMFD) is employed to accelerate the calculation. A pin cell is chosen as the coarse mesh. The CMFD solution provides the MOC with much faster converged fission and scattering source distributions. The LOTUS code is developed and the numerical results show that the code is precise for engineering application and the CMFD acceleration is effective.
Study on Improvement of Analytic Depletion Calculation Method
Authors: Huang Kai , Wu Hongchun , Cao Liangzhi , Zheng Youqi
Published: February 2014

The transmutation trajectory analysis (TTA) method is a traditional analytic depletion calculation algorithm. It’s capable of providing very accurate solutions with cutoff value being sufficiently small, but the main drawback is expensive time consumption. In this paper, the depletion problem was rephrased as a directed-graph module with additional properties attributed to vertices and edges. Based on this model a new approach for TTA implementation was conducted, to which dynamic programming technique could be introduced easily. Besides it, a new analytic algorithm is proposed based on the above model, which gives each nuclide a time dependent concentration function instead of a scalar value that other algorithms could only offer. One of the advantages brought by this algorithm is making time integration value calculations straightforward. Both of these two approaches are programmed. The resulting TTA code was proved to be very efficient, the running time for a typical ORIGEN2 problem with cutoff equals to 10−15 could be of the same magnitude with time spent on library reading process. The codes are analyzed in terms of time and space complexities, which could offer a theoretical point of view on their behaves.
Research on High and Low Pressure Filter Cartridge Replacement and Transport Process and Related Operating Equipments Design for AP1000 Nuclear Power Project
Authors: You Xiangfei , Chen Kang , Xing Jiansheng , Shen Haida , Meng Lei , Shi Ke
Published: February 2014

Based on the brief introduction of AP1000 Nuclear Power Project high & low pressure filter and filter cartridge location information, and through researching and analyzing the whole operation process of taking out spent filter cartridge from filter vessel, transferring the spent filter cartridge to spent filter storage area and installing new filter cartridge, the reasonable and feasible filter cartridge replacement and transport operation process was proposed. And based on the proposed operation process, the related equipments design of filter cartridge replacement and transport were optimized. Reasonable and feasible operation process combining with perfectly shielding related equipments can maximally reduce radioactive occupational exposure risk to operating personnel and effectively prevent operating equipments and the ambient environment from radioactive pollution. During filter cartridge replacement and transport, the maximum radiation dose rate at operating position (cared position) was reduced to 0.830mSv/h by radiation shielding calculation. Meanwhile, the research content of this paper can also provide technical reference for nuclear power plant operators to actually operate filter cartridge replacement and transport. This paper also proposed the improving suggestion for follow-up projects on the basis of Self-Reliance Supporting Project existing problems on filter cartridge replacement and transport, and briefly summarized the innovation and application value of this paper.
Primary Shielding Design for an Optimized Molten Salt Reactor
Authors: Zhang Zhihong , Xia Xiaobin , Wang Jianhua , Li Changyuan
Published: February 2014

Molten salt reactor (MSR) system, a candidate of the Generation IV reactors, has inherent safety, on-line refueling and good neutron economy as typical advantages.

An optimized MSR is developed by changing the size of fuel channel and the graphite-to-molten salt volume radio, based on the Molten-Salt Reactor Experiment (MSRE), which was originally developed at the Oak Ridge National Laboratory (ORNL). In this paper, shielding calculations for the optimized MSR are presented. The goal of this study is to determine the necessary shielding to decrease the neutron and gamma dose rate to the acceptable level according to national regulations. The operating temperature of the optimized MSR is designed in the range of 500 °C–700 °C, heat removal is also considered in the shielding design.

The shielding calculations are carried out by using Monte Carlo method. The shielding system of the optimized MSR consists of 7 zones: the core, the core can, the reactor vessel, the thermal shield, the reactor cell containment, the shield tank and the concrete wall. The combinations of shielding materials in the thermal shield were evaluated. The thermal shield filled with carbon steel balls and circulating water gets an excellent shielding performance and heat removing effects. The neutron spectra and dose distributions, as well as the energy deposition over different shields have been analyzed. The total neutron dose rate outside the thermal shield is attenuated by a factor of about 104, and the gamma dose rate by a factor of about 103. These results show that the shielding design could low dose rate to an acceptable level outside the shielding and far below dose limit required.
Seismic Response Spectrum Analysis on Advanced PWR Coolant System
Authors: Duan Rong , Tong Lili , Cao Xuewu
Published: February 2014

The pressure boundary of reactor coolant system (RCS) consists of reactor vessel, pressurizer, steam generators, coolant pumps, main pipes and other crucial equipments. During operating, the RCS and its equipments withstand mechanical loads, pressure fluctuations, even earthquakes and other extreme loads, which shows that the vibration characteristics and seismic response are significant for the safe operation. In this paper, the RCS of advanced passive pressurized water reactor (PWR) is studied. The three-dimension finite element model, consisting of one reactor vessel, pressurizer, two steam generators and four coolant pumps, is constructed to simulate the seismic response. Based on modal analysis, the deformation and stress response of RCS under Safety Shutdown Earthquake (SSE) loads is analyzed under three orthogonal directions with multi-angle. In each case, the two horizontal input spectra are rotated on z axis with 30 degrees. The variation of seismic response with spectrum input angle shows the best and worst orientation for anti-earthquake and provides basis for the design, installation and improvement of RCS. Moreover, compared with static analysis of RCS, the stress under extreme condition is linearized and assessed with ASME standards. The result shows that all the values satisfy the stress criteria and the structure and material of the RCS meet the design requirements. More work on the parameters sensitivity analysis will be done in further study.
Analysis of Automatic Depressurization System Stage 4 Valves Inadvertent Actuation Scenario for a Certain Passive Power Plant
Authors: Lu Lu , Yu Jianhui
Published: February 2014

The automatic depressurization system (ADS) stage 4 valves inadvertent actuation scenario is one of the risk-significant cases in probabilistic risk assessment (PRA) analysis of passive power plant. Based on mini-ASTRUM (automated statistical treatment of uncertainty method) and CQD (code qualification document) method together, the calculation which utilizing WCOBRA/TRAC and HOTSPOT codes has been made in this paper. And the results indicate that, the calculated peak cladding temperature (PCT), local maximum oxidation (LMO), and core-wide oxidation (CWO) are all low enough, which meet the three design acceptance criteria defined in 10 CFR 50.46 emergency core cooling system (ECCS).
CFD Analysis of Thermal Stratification of China Lead Alloy Cooled Research Reactor (CLEAR-I)
Authors: Zhao Pengcheng , Chen Zhao , Zhou Tao , Chen Hongli
Published: February 2014

Thermal stratification phenomena is observed in the upper plenum under the reactor scram transient of CLEAR-I. This will bring a bad effect on the natural circulation of the coolant Lead bismuth eutectic (LBE) in the primary circuit, which give rise to thermal stress on structural components. A CFD analysis of 2D axisymmetric domain to predict thermal stratification in scram conditions from full power operation has been performed by commercial code ANSYS 14.0. From the result, it can be seen that the thermal stratification can be formed in the hot pool under reactor scram conditions, the interface of the clear stratification was located below the inlet window of heat exchanger, but it will never stop the natural circulation of the primary circuit. The results obtained will be used to make an optimized design of China Lead Alloy Cooled Research Reactor (CLEAR-I).
Thermal-Hydraulic Simulations of Guard Containment in Depressurization Accidents of a Gas-Cooled Fast Reactor
Authors: Xu Zhanjie , Jordan Thomas
Published: February 2014

A gas-cooled fast reactor is designed as an advanced nuclear reactor in next generation in the EU. In depressurization accident scenarios, pressurization caused by a release of helium from the primary system with a higher pressure into the guard containment would endanger the integrity of the containment. In the design stage, the released source term is analyzed theoretically, and is applied as a boundary condition in the 3D CFD code simulation to the transient pressurization process. The simulation results supply a reference value about the design pressure of the containment.
Effect of Initial and Boundary Conditions on the Formation of Reactor Vessel Pressurized Thermal Shock
Authors: Vyshemirskyi Maksym , Mazurok Aleksander , Nosovskyy Anatoliy
Published: February 2014

The effect of initial and boundary conditions on the reactor vessel Pressurized Thermal Shock (PTS) was analyzed. For example unit 1 of South-Ukrainian Nuclear Power Plant (NPP) with WWER-1000/V-302 reactor was selected. Thermal-hydraulic analysis was made in accordance with methodology of Reactor Pressure Vessel (RPV) brittle fracture analysis by using RELAP5/mod3.2 code. Also, a detailed model of downcomer, which provides a realistic behavior of the fluid, including the mutual mixing of flows with different temperatures was used for these calculations. As an Initial Event (IE) the large secondary leak, namely Steam Generator (SG) Main Steam Line (MSL) break, was considered. According to the International Atomic Energy Agency (IAEA) recommendations all calculations were performed under conservative approach. Also, the asymmetric overcooling was considered. As a result of a series of thermal-hydraulic calculations the most conservative scenario for a group of IE associated with the secondary leaks was obtained.
Investigation of Thermophysical and Nuclear Properties of Prospective Coolants for Generation-IV Nuclear Reactors
Authors: Dragunov Alexey , Saltanov Eugene , Pioro Igor , Ikeda Brian , Miletic Marija , Zvorykina Anastasiia
Published: February 2014

Recently, a group of countries has initiated an international collaboration, the Generation IV International Forum (GIF), to develop the next-generation nuclear reactors. The GIF program has narrowed the design options of nuclear reactors to the following six concepts:

1) SuperCritical-Water-cooled Reactor (SCWR);

2) Sodium-cooled Fast Reactor (SFR);

3) Lead-cooled Fast Reactor (LFR);

4) Molten Salt Reactor (MSR);

5) Gas-cooled Fast Reactor (GFR); and

6) Very-High-Temperature Reactor (VHTR);

The purpose of this paper is to compare main thermophysical, corrosion, and neutronic properties of the Generation-IV reactors’ coolants within the proposed range of operation.
Circularity Error Analysis of Axial Displacement Sensor Detecting Surface for Active Magnetic Bearings of 10MW High Temperature Gas-Cooled Reactor (HTR-10GT)
Authors: Zhao Jingjing , Zhou Yan , Sun Zhe , Yu Suyuan
Published: February 2014

The 10MW high temperature gas-cooled test module reactor (HTR-10GT) with the core made of spherical fuel elements was designed and constructed by the Institute of Nuclear and New Energy Technology of Tsinghua University in China. In the HTR-10GT, turbo-compressor and generator rotors are connected by a flexible coupling and a decelerator. The rotors, restricted by actual instruments and working environment, must be supported without any contact and lubrication. Active magnetic bearing (AMB), is the best way to suspend and stabilize the position of rotors of HTR-10GT. In AMB system, the displacement sensors are designed for converting measured physical value into proportional electric signals and transferring it into the control and monitoring system. The characteristics and influence of circularity error of the axial displacement sensor detecting surface is analyzed according to AMB with five degrees of freedom. In order to reduce the circularity error of the magnetic bearing system, the sensor detecting surface requires high-precision machining.
Theoretical and Experimental Study on Natural Circulation Relevant to Nuclear Heating Reactor
Authors: Wang Dingqu , Jia Haijun , Sun Yanfei , Liu Yang
Published: February 2014

This paper presents a one-dimensional theoretical model for describing the steady state flow and heat transfer processes of single-phase natural circulation flow. By mean of experimental investigation, the relationship of flow resistance and flow rate is obtained, and then utilized to acquire an accurate computation model. The response behaviors of thermal-hydraulic parameters varied with the heating power are also observed. The expression of resistance and flow rate is attained as ΔP ∝ G1.40, and the relationship between flow rate and heating power is G ∝ Q0.38. Experimental results show good agreements with the theoretical model. Moreover, the influence of flow resistance in the core inlet is studied.
CFD Analysis on Hydrogen Risk in Subcompartment of the Containment
Authors: Zou Zhiqiang , Deng Jian , Zhang Yu
Published: February 2014

Hydrogen combustion in the containment building may be a threat to containment integrity. Especially in some subcompartments, local hydrogen detonation likely to happen, which may destroy the containment structure or the safety equipments in the containment. A study has been carried out using the GASFLOW three-dimensional CFD code to evaluate the hydrogen distribution in the subcompartment during a severe accident. The GASFLOW calculation has provided detailed results for the spatial distribution of gas concentrations in the reactor coolant pump (RCP) compartment and the steam generator (SG) compartment within containment, the high local hydrogen concentration was found in the local area. Then, three kinds of different hydrogen mitigation measures which were optimizing compartment structure, hydrogen recombiner and igniter were analyzed and compared. The σ criterion and λ criterion are used for conservative estimates of the flame acceleration (FA) potential and the possibility of deflagration to detonation transition (DDT) in the compartment after using the mitigation measures, respectively.
Performance Analysis of a Novel PXS With Organic Rankine Cycle (ORC) System
Authors: Zhang Shengjun
Published: February 2014

With the increasing of core thermal power of the nuclear power plant, the decay heat of the core increases in the accident. Therefore, the heat removal capacity of the PXS should be enhanced to fulfill the requirement of core safety.

A new scheme is put forward to improve the cooling capacity of PXS and provides long-term power for station blackout (SBO) accident or loss of normal feedwater. In this system, the Organic Rankine Cycle is incorporated between the hot leg and cold leg of PRHR. The decay heat of the core is the heat source and the cooling pool outside the containment is the cool source. The natural circulation of the primary loop is established due to the density difference. The primary fluid flows into the evaporator of the ORC system, where the working fluid of the ORC system is evaporated. Then the temperature of the primary fluid is decreased. The vaporized working fluid drives the expander, which is coaxially fixed with the fluid pump, to generate the power. Finally, the exhausted vapor flows into the condenser and the residual heat is discharged outside of the containment simultaneously. The working fluid in the condenser is pumped into the evaporator by the fluid pump for liquid supplement and the cycle keeps on working continuously.

A steady state analysis is performed on a 1700MWe nuclear power plant with ORC as the heat removal system. The heat transfer area of the ORC evaporator is fixed as 487.7m2, which is the same as the area of PRHR HX. The efficiencies of fluid pump and expander of ORC system are assumed as 0.75 and 0.8, respectively. The decay heat of the core is about 67.62MWe, which is 1.38% of the core full power. The working fluids are screened and R141b offers excellent performance. The efficiency of fluid pump and expander are assumed as 0.75 and 0.8, respectively. The condensing temperature is assumed as 80°C and the evaporating temperature is 160°C. The results show that 7.83MWe will be generated by the ORC system and the heat transfer area of the condenser is about 994.5m2. The residual heat of 59.79MWe will be discharged to the water tank outside the containment.
Thermal Hydraulic Analysis of a SCWR Core Concept
Authors: Zhang Weizhong
Published: February 2014

A Super Critical Water Reactor (SCWR) single channel analysis model was built considering energy coupling between water rods and coolant channels. By using this model, a comparative study was performed on SCWR fuel assembly arrangement such as the size and location of water rods and fuel pins. Cladding temperature, pellet centerline temperature, and coolant core-outlet velocity were selected as comparison criteria. Preliminary core conceptual design of 1000 MWe SCWR was carried out and two sets of main parameters for SCWR core were proposed.
Identification of Flexible Rotor Suspended by Magnetic Bearings
Authors: Sun Zhe , Zhao Jingjing , Shi Zhengang , Yu Suyuan
Published: February 2014

Magnetic bearings are widely applied in High Temperature Gas-cooled Reactor (HTGR), where the rotating machineries are running under high purely helium environment. In designing and adjusting a magnetic bearing system, the mathematical model of the rotor plays an important role. Identification is a useful method to obtain the model of a rotor. However, there are some practical difficulties of identifying a magnetic bearing-rotor system without force sensors. This paper proposes an identification method for flexible rotor suspended by magnetic bearings. In this method, two experiments under different bearing stiffness are performed, the models obtained by these two experiments are then transformed to the desired rotor model and the influence of bearing stiffness is eliminated in this transformation. The proposed method is validated on an experimental system with a five degree-of-freedom suspended flexible rotor.
Based on the Meta Analysis Thought of Nuclear Power Equipment Maintenance Strategy Optimization
Authors: Li Shuqing , Mou Huaming , Xiao Hongxing
Published: February 2014

Nuclear power equipment safety and reliable operation has important social meaning and economic significance, especially some scale is huge, and the mechanism is complex, the impact of decommissioning major and the nuclear safety and to protect public health plays an important role in the equipment, to ensure their safety and reliable operation has an important meaning and the higher maintenance demands.

At the same time, the nuclear power equipment maintenance has maintenance field condition multifarious, difficult, short construction period, quality, safety and technology requirements higher characteristic, so to meet maintenance objectives, the nuclear power plant equipment especially the key and important equipment, the urgent need to adopt more advanced, and more scientific strategy for repair.

Meta analysis method is a more popular on the same theme of multiple independent test results are comprehensive statistical method. At present, it is a very effective comprehensive analysis method, and has already developed mature in many domestic and foreign industries and fields.

This paper based on the Meta analysis thought optimization equipment maintenance strategy, working to improve nuclear power equipment especially the key and important equipment maintenance more scientific, reliability and security.
Simulation of the Atmospheric Dispersion of Radionuclides Using Gaussian Plume Model
Authors: Yang Ye , Cao Bo , Chen Yixue
Published: February 2014

The Chernobyl accident and Fukushima 1 Nuclear Power Plant accident are the most serious accidents in the history of the nuclear technology and industry. A large amount of radioactive materials from nuclear power plant were released, leading to huge damage and long-term effect on the environment as well as the human health neighbor to the plant. Therefore, simulating the transport and transformation of radionuclides in the atmosphere is significant for decision makers to take steps at all level. Now, many different dispersion models are widely applied and used to simulate the transport and transformation of radionuclide such as Gaussian model, Lagrangian model and Eulerian model. Though the Eulerian or Lagrangian models have several advantages, such as high spatial resolution, fully 3D descriptions of the meteorological, the simple Gaussian plume model is still widely chosen because of its higher accuracy and faster calculation. In this study, the atmospheric dispersion of leaked radioactive material during nuclear accident is simulated by using Gaussian plume model. The relative concentration distribution of the radionuclides and the trajectory of the distribution centrode are obtained in taking account of different geographical environments, wind direction, wind velocity, and stability category. These results can provide a favorable evidence for the management of nuclear accident emergency.
Analysis and Applications of a Two-Fluid Multi-Field Hydrodynamic Model for Churn-Turbulent Flows
Authors: Montoya Gustavo , Liao Yixiang , Lucas Dirk , Krepper Eckhard
Published: February 2014

Today Computational Fluid Dynamic (CFD) codes are widely used for industrial applications, mostly in the case of single phase flows in automotive or aircraft engineering, but multiphase flow modeling had gain an increasing importance in the last years. Safety analyses on nuclear power plants require reliable prediction on steam-water flows in case of different accident scenarios. This is particularly true for passive safety systems such as the GEKO component of the KERENA reactor. Here flashing may occur in the riser (Leyer and Wich, 2012). In such case, high gas volume fractions and the churn-turbulent flow regime may ensue. In the past, the codes for the prediction of churn-regime have not shown a very promising behavior. In this paper, a two-fluid multi-field hydrodynamic model has been developed based in the Euler-Euler framework. The main emphasis of this work has been on the modeling and applicability of various interfacial forces between dispersed gaseous phases and the continuous liquid, as well as bubble-bubble interactions, and the evolution of different bubble sizes in an adiabatic vertical pipe inside the churn-turbulent flow regime. All the expected mechanistic models that intervene in this flow pattern have been taken into account including drag force, wall force, lift force, turbulent dispersion, and bubble induced turbulence. Bubble breakup and coalescence has been defined (Liao et al., 2011), and in order to design a polydispersed model related to reality, the inhomogeneous MUSIG approach (Krepper et al., 2008) has been used to defined an adequate number of bubble size fractions which are arranged into different groups with their own velocity field. Based on these models, a series of simulations were made on the framework of ANSYS CFX 14.0, and all of the calculations were further validated with experimental data extracted from the TOPFLOW facility at the Helmholtz-Zentrum Dresden-Rossendorf. Different water and gas flow rates were used inside the churn-turbulent flow regime, as well as for the transition from bubbly to churn flow. The calculated cross-section averaged bubble size distributions, gas velocities, and time averaged radial profile for the gas fraction have shown a promising agreement with the experimental data. Nevertheless there are also clear deviations which indicate shortcomings of the present modelling. In order to further improve the modeling of this flow regime, a discussion based on the results will be used to shown a series of limitations of the actual modeling and possible solutions to be implemented in future works.
Numerical Simulation on 3D Flow in the Fuel Transfer Canal and Local Flow Field Analysis
Authors: Guo Qiang , Wang Hui
Published: February 2014

By application of CFD codes, 3D complex flow field and temperature field in the transfer canal are simulated. It indicates that there is obvious cross-flow in the complex flow path of spent fuel container, and the natural circulation flow brings an inhomogeneous heat-transfer condition to the spent fuel assembly. A model of typical unit of canal is built, and local flow field and temperature field are analyzed with CFD code. The results of thermal hydraulics analysis could be a good reference for nuclear safety research on spent fuel assembly. Considering a conservative operating condition, a case with conservative boundary conditions is investigated by CFD code with 3D model. The result shows that fuel rods in the local region, which are in the container upside and far from holes of wall and roof, get the worst cooling effect, and even maybe boiling occurs in clad surface. Based on the thermal-hydraulics analysis, optimized mechanical designs of container are suggested, the new design would optimize the local flow field, and increase the flow velocity of water, to construct more unhindered convection there, as a result, make spent fuel rods in the bad cooling region get better cooling.
Fracture Mechanic at Design Level for New Reactors
Authors: Faidy Claude
Published: February 2014

For new plants in particular, or for some replaced components a new failure mode has to be considered: flaw tolerance and rupture.

Different cracks have to be considered at the early stage in design in different locations with different crack sizes and crack locations, in particular for components in the brittle or transition temperature range. In addition, for some location thermal ageing or radiation ageing have to be considered.

Potential fatigue or corrosion crack growth has to be also considered.

Another aspect that needs to consider fracture mechanic is the Leak Before Break justification or the Break Exclusion justification of some components or piping systems.

This paper will summarize how these requirements are considered in French Codes RCC-M, RCC-MRx and RSE-M for normal, upset, emergency, faulted, severe accident and test conditions.

On-going developments, like Appendix G modernization, will be presented to conclude the paper.
Licensing Considerations of a Fluoride Salt Cooled High Temperature Test Reactor
Authors: Xiao Yao , Hu Lin-wen , Forsberg Charles , Qiu Suizheng , Su Guanghui
Published: February 2014

The Fluoride salt cooled High temperature Reactor (FHR) is an innovative reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride salt cooled High temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal hydraulic licensing analysis of an FHTR using SINAP’s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting Safety System Settings (LSSS) considering uncertainties for forced convection operation are obtained. The LSSS power and coolant outlet temperature are 24.6 MW and 720 °C, respectively.
Fundamental Study to Evaluate Mechanical Property Change Associated to Dislocation Behavior in Irradiated Austenitic Stainless Steels by Incorporating Thermal Fluctuations
Authors: Asari Keisuke , Miyashiro Satoshi , Itakura Mitsuhiro , Okita Taira
Published: February 2014

Molecular dynamics simulations were conducted with three interatomic potentials for face-centered cubic (FCC) metals with different stacking fault energies (SFEs), and the effect of the SFE on the dislocation behaviors was evaluated by incorporating thermal fluctuation. The separation distance between two partial dislocations fluctuated and was changed locally by thermal vibration, and the fluctuation increased with temperature. For a low-SFE metal, thermal vibration can hardly induce reversions into a perfect dislocation. This causes a characteristic morphology for interactions between a line dislocation and irradiation-induced defects for austenitic stainless steels.
Preliminary Study on the Feasibility of Utilizing the Thermal Fissile Breeding Capability of the Th-U Fuel Cycle in HTR-PM
Authors: Xia Bing , Li Fu
Published: February 2014

HTR-PM is a demonstration plant of the modular high temperature reactor with two pebble-bed cores of 250 MWth. Since the early HTRs, such as AVR and THTR in Germany, thorium-based fuel has been regarded as an important fueling option. In this work, the feasibility of maximizing thorium utilization and minimizing the refueling effort of uranium fissile under the framework of HTR-PM is investigated. The preliminary neutronics features of the Th-233U fuel cycle in the equilibrium state of the HTR-PM are analyzed. Two types of fuel loading schemes are considered: the ThU-MOX scheme, namely the so-called “ThOX” fuel in literatures, and the SEP scheme, which means “separate” fuel pebbles loaded with thorium oxide and uranium oxide, respectively. The ThU-MOX scheme utilizes the mixed Th+HEU oxide fuel particles in all the fuel pebbles in the core, in which the enrichment of HEU is 93%. The SEP scheme utilizes the separate thorium pebbles and LEU pebbles mixed homogeneously in the core. The preliminary results on the ThU-MOX scheme indicate that thorium mixed with HEU in the fuel particle can lower the 235U loading requirement per energy generated, compared with the normal LEU loading scheme of the HTR-PM, and enhance the safety performance for high heavy metal loading cases. On the other hand, the results of SEP scheme reveal that the self shielding effect of the thorium particles depresses the absorption of thorium and the utilization of 233U. However, the situation can be improved by lengthening the residence time of the thorium pebbles. Furthermore, more realistic features are investigated as the basis of future works, including the initial core and the running-in phase, the impact of control poisons.