bluuebunny/crossref_metadata_2025 huggingface dataset:
New Method of Coolant Mixing Studies at the Operating WWER-1000 Units
Authors: Ponomarenko Grigory L. , Berkovich Vadim Ja. , Bykov Mikhail A.
Published: February 2014

Paper [1] describes a unique method and basic results of coolant mixing experimental study, carried out on the working power unit WWER-1000 of “Bushehr” NPP. Method is characterized by the use of an emergency boron insertion system for creating the nonuniform distribution of indicator (tracer), and also by the use of all regular systems of neutron and temperature monitoring for increasing the authenticity. Method showed high reliability, effectiveness, availability and flexibility. The corresponding application for the invention of a new method is applied for a patent. Present work supports the conclusions of the paper [1] and contains the more detailed additional information and results according to the mixing indices, and also to the dissipation of temperature indicator in different sections of circulation loop. Some results of post-test modeling by calculation code KORSAR/GP are presented too. Spatial and temporal detailing relate also to the dynamics of appearance, stabilization, propagation and disappearance of boron and boron-free “spots” in the core with start-up and trip of the emergency boron insertion pumps.
Corrosion Behavior of Zircaloy-4 in Methanol Solution at 320 °C Under Gamma-Irradiation
Authors: Domae Masafumi , Hojo Kosho , Sugino Wataru
Published: February 2014

It has been pointed out that high concentration dissolved hydrogen is one of the important factors of PWSCC (primary water stress corrosion cracking) in the primary systems of pressurized water reactors. Application of a substitution for hydrogen may be a fundamental countermeasure of PWSCC. The authors are developing a new water chemistry technology of a hydrogen alternative to suppress PWSCC. In the present paper, corrosion tests of Zircaloy-4 were performed in deaerated 5 mmol dm−3 methanol solution at 320 °C in the absence and presence of gamma-irradiation. The nominal absorbed dose of the test water was 100 kGy. After the immersion tests, the specimens were analyzed. Weight gain per unit surface area, thickness of oxide film and hydrogen storage were measured. In addition, Raman spectroscopy was carried out, to investigate possible deposition of organic compounds on surface of the specimens. The corrosion behavior of Zircaloy-4 without irradiation agreed with literature data. It was concluded that the presence of methanol did not affect the corrosion behavior of Zircaloy-4. The corrosion behavior of Zircaloy-4 hardly depended on 100 kGy gamma-irradiation. On the Raman spectra of the specimens after the immersion tests, the Raman peaks ascribed to polyethylene or graphite were not found. The deposit of decomposition products of methanol would be negligible if any. It seems that polymerization is not the major process in thermal decomposition and radiolysis of methanol, but methanol decomposes into CO2 or carboxylic acids.
Analysis of Advanced PWR Loading Schemes for Transuranic Incineration in Thorium
Authors: Lindley Benjamin A. , Zainuddin N. Zara , Franceschini Fausto , Parks Geoffrey T.
Published: February 2014

It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favorable impact on the MTC and void coefficient fostered by use of thorium (Th), the possibility of performing Th-TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. The simplest way to reduce the moderation in a PWR is to increase the fuel pin diameter. This configuration improves the trade-off between achievable burn-up and MTC, but is ultimately limited by thermal-hydraulic constraints. Heterogeneous recycle with the bred uranium (U3) and the TRU are arranged in separate pins was found to be neutronically preferable to a homogeneous configuration. Spatial separation also enables the U3 and TRU to be refueled on different batch schemes. These techniques allow satisfactory discharge burn-up while ensuring negative MTC and fully voided reactivity, with the pin diameter of a standard PWR increased from 9.5 mm to 11 mm. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron. It seems necessary to control the core using control rods to keep the fully voided reactivity negative. A preliminary analysis indicates that this is feasible.
Multiphase Flow Simulation of Subcooled Boiling Using the N-Phase Approach in TransAT
Authors: Thomas S. , Narayanan C. , Lakehal D. , Gong Y.
Published: February 2014

In Pressurized Water Reactors (PWR), convective boiling occurs at the heated walls, which are superheated while liquid bulk is subcooled at a given operating pressure. The physical modeling of the phenomenon is complex, as it requires consideration of phase change and turbulence. The combinations of Reynolds-Averaged Navier Stokes (RANS) and wall boiling models have found moderate success in the past. Succesful modeling depends on the nature of the problem and its operating conditions, the manner of the model implementation and the quality of the computational platform. In the present work recent developments executed in the TransAT code for subcooled boiling are demonstrated. To better capture the phase change, the modified mixture approach is employed where the temperature of the N-phases are resolved separately. This method has been more suitable for phase-change problems, when compared to a single mixture temperature. To check the validity of the modeling, results are evaluated against the DEBORA experiments carried out at CEA Grenoble.
Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG
Authors: Yan Jinquan , He Yinbiao , Li Gang , Yu Hao
Published: February 2014

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.
Safety Culture Promoting Activities After Fukushima Accident
Authors: Kurata Satoshi , Suzuki Tetsuya , Sekine Yoshihiro
Published: February 2014

Nuclear safety culture has been one of the most important aspects for safe and reliable operations of nuclear power plants worldwide.

Promoting nuclear safety culture has been a very important task for Japanese nuclear operators since 2007, when it was introduced into the Safety Regulations.

CEPCO has positively promoted safety culture. Although TEPCO also promoted safety culture, the Fukushima accident occurred in March 2011. The official investigation reports on the accident pointed out that one cause of accident could be safety culture problems. Therefore we should seriously consider effectiveness of safety culture promoting activities which we worked on.

This study presents issues and appropriate improvements for the safety culture promoting activities reflecting the Fukushima accident and our experiences in this five years.
Effect of Feeder Inner Structure on Feeding Performance for Absorber Sphere Pneumatic Conveying
Authors: Li Tianjin , Chen Feng , Zhang He , Yang Hairui , Huang Zhiyong , Bo Hanliang , Yan He , Qi WeiWei
Published: February 2014

The sphere feeder is an important component of the absorber sphere shutdown system for absorber sphere conveying. Absorber sphere is entrained and fluidized in the feeder by conveying gas to form gas-solid flow. The effect of feeder inner structure, i.e. the draft tube height, on the feeding performance was investigated. Experiments were conducted in the positive pressure conveying system with ambient air as the source gas. The glass sphere was used to replace the absorber sphere. Three types of feeder inner draft tube relative position were considered. Five combinations of the draft tube height were tested with feeder inlet air velocity in the range of 15∼30 m/s. Experimental results showed that feeder inner draft tube height played an important role on the feeding performance. Different characteristics for the starting process of the sphere conveying were observed. The difference of sphere blown away pressure and steady conveying pressure could be reduced to about 5 kPa, which was negligible from the view point of conveying blower selection in engineering application. Sphere stagnation region was observed near the bottom of the feeder. The obvious boundary of stagnated sphere and fluidized sphere was formed for the steady conveying. Two types of control model were observed for sphere steady feeding rate. One was that the sphere steady feeding rate was determined by sphere flow through the orifice. The other was that the sphere steady feeding rate was determined by gas-sphere fluidization and entrainment. These results were important for optimization sphere feeder design.
Study on Network Radiation Environmental Monitoring System
Authors: Zhang Huai Qiang , Liu Qing Cheng , Liu Yu Juan
Published: February 2014

With the application of nuclear technology in many fields and the development of nuclear power, the people pay more and more attention to which the radioactive substances are harmful to the human, environmental pollution and nuclear security. In order to prevent the dangers of radioactive substances on the professional staff and the public, the pollution of the environment, it is particularly important to establish a scientific and perfect radiation environment monitoring system. In this paper, the radiation environmental monitoring system based on network is designed for environment radiation situation of the original enterprise, such as the uranium mining, nuclear power plant, uranium hydrometallurgical plant, nuclear fuel processing plant. The network technology is introduced in the monitoring system which links the field monitoring system, data center service system, data transmission system and client system. the worker can control the on-site monitoring points, data acquisition and transmission in any of the network interface functions, complete displaying, analyzing and processing in the remote server, meanwhile the network power supply technology is applied in the monitoring system which can supply electricity to on-site monitoring points. According to the test, the radiation environmental monitoring system which based on network technology is easy to apply and maintain, with a certain degree of practicality and economy.
Research on the Stainless Steel Tubes and Pipes Material Standards of Demonstration Fast Reactor Power Plant
Authors: Li Yaping , Song Guangdong
Published: February 2014

The main characteristics of the sodium pipe system in Demonstration Fast Reactor Power Plant (DFRPP) are high-temperature, thin-wall and big-caliber, which is different from high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term operate in the environment of liquid metal sodium. How to guarantee the reliability of materials in high temperature are most important in material option.

Engineering design depend on the criterion. Material standards are different in different countries, and corresponding construction codes are different too. Comparing the stainless steel pipe material standers at home and abroad and analyzing the material standards’ difference according to different construction codes, a stainless steel pipe material criterion system is put forward in this paper which is applicable for the DFRPP.
Development of PCCSAP-3D Code for Passive Containment: Models of Noncondensable Gases, Aerosols and Fission Products
Authors: Li Ran , Yu Jiyang
Published: February 2014

PCCSAC-3D is a code originally developed for AC600 containment thermo-hydraulic analysis. Its validated capabilities include simulating the behaviors of steam-air mixture and liquid water under the unique conditions of an AC600/AP1000 containment after a DBA. The film-tracking model applied gives it the ability to simulate the liquid film both outside and inside the steel containment. Refined with some new models, the new version of the code, named PCCSAP-3D, can cover hydrogen behavior, fission products behavior (in the form of gas and aerosol) and iodine behavior.

In the module of noncondensable gas, diffusion of up to 11 species are taken into consideration. A user-definable recombiner/ignitor model is developed to accommodate different types of hydrogen recombiners and ignitors.

Given the source term as a boundary condition, the fission products model would be able to track up to 64 radio-isotopes after a LOCA. The leakage and spontaneous decay is accounted for all of these nuclides. Besides, the noble gases, gaseous iodine and fission product aerosols are treated separately. There is no removal mechanism of noble gases. Whereas removal mechanisms of radio-aerosols considered include spray, gravitational sedimentation, diffusio-phoresis and thermo-phoresis. A simple model for gaseous iodine comprises organic iodine and elemental iodine, in which the effects of spray and liquid adsorption are treated integrally.

To evaluate the radioactivity consequences of a certain accident, a radioactivity calculation model is brought out to convert the molar concentration or mass concentration of radioactive material into radioactivity concentration.

The new version of PCCSAP-3D code with models aforementioned is preliminarily validated by comparing the simulation results with safety analysis results reported in AP1000 Design Control Document. The accident scenario is set as a design basic LOCA with core melt.
Numerical Research on Performance of AP1000 Passive Containment Cooling System Without Falling Film
Authors: Wang Hui , Guo Qiang , Chen Qiaoyan
Published: February 2014

Passive Containment Cooling System (PCS) is one of the most important passive safety features in AP1000 nuclear power plant. Numerical Simulation of PCS without falling film was conducted using CFD method in this paper to investigate heat transfer performance of PCS. The results indicate that: at the time of 72 hours after break accident, there is no water in passive containment cooling water storage tank (PCCWST), PCS cannot completely remove reactor core decay heat to environment just with the effect of natural convection and thermal radiation; mass flow rate of natural convection and PCS heat removal power increase with the increase of the outside surface temperature of steel containment, but the increasing amplitude of the former is smaller than the latter significantly, and the fraction of thermal radiation in the PCS heat removal power is the smallest when the outside surface temperature of steel containment is 80 °C; mass flow rate of natural convection and PCS heat removal power decrease with the increase of environmental temperature, while the fraction of thermal radiation in the whole heat removal power becomes bigger with the increase of environmental temperature.
The Experimental Consideration for TBM Mock-Up Effect on Plasma Performance Based on MAPES Platform in EAST
Authors: Liu Songlin , Ding Fang , Chen Xiangcun , Pu Yong , Li Jia , Ma Xuebin , Luo Guangnan
Published: February 2014

EAST can provide better opportunities to contribute development of ITER-relevant plasma physics and engineering because it has ITER-like configuration, and has achieved 10s H-mode plasma, and aims steady-state operation of DD high performance plasma. The impact of Test blanket module (TBM) using RAFM (reduced activation ferritic/martensitic) steels on tokomak plasma is a major concern in ITER operations. In order to assess this effect due to TBM local ripple, an experiment plan of TBM mockup using RAFM steel is being planned on MAPES (Material and Plasma Evaluation System) in EAST. This paper reports experimental consideration on MAPES based on magnetic analysis and ripple calculation at separatrix point. The relevant experiments strategy and plan in EAST are also proposed.
Pressure Drop Characteristics of Flow Boiling in Narrow Rectangular Channel
Authors: Lv Shu-ping , Pan Liang-ming , Chen De-qi
Published: February 2014

Pressure drop is the earliest and the most extensive issue in the research of two phase flow. Due to the influence of various factors on two phase flow pressure drop, there is no accurate and general correlation on this issue yet. In this paper, pressure drop characteristics of flow boiling were studied in a single vertical narrow rectangular channel with cross-section of 47 mm by 1.8 mm using water as working fluid. Subcooled flow boiling was experimentally investigated. The experimental conditions included inlet subcooling (2°C∼18°C), mass flow rate (222.3∼457.4 kg.m2s−1) and heat flux (113.6∼270.5 kW/m2). The influence of these parameters on the characteristics of flow boiling pressure drop has been investigated. The experimental results show that the conventional correlations failed to predict the two-phase flow pressure drop in this narrow rectangular channel. According to experiment results, a modified Chisholm correlation was proposed, and the new correlation gave more accurate predictions of the pressure drop.
Primary Simulation: The Yield of 14C Produced in PWR During Normal Operation
Authors: Liu Shuhuan , Zhang Xin , Wang Sanbing , Zhao Yaolin , He Chaohui , Wu Yuelei , Chen Wei
Published: February 2014

The mechanisms of the radionuclide 14C production and releasing from pressurized water reactor (PWR) nuclear power plant (NPP) are simply described in the paper. The yield calculation model for 14C produced in PWR is offered in this work. The averaged neutron flux in reactor core under normal operation condition is simulated with MCNP5 according to the simplified Daya Bay NPP reactor core structure and other parameters, the yields and activities of 14C in different parts of the reactor core including fuel, fuel cladding, nitrogen impurities in fuel and coolant are calculated and compared with Origen2. The simulation results showed that the yields of 14C produced in different parts change linearly with operation time. The rations of 14C introduced by nuclear reactions in coolant, fuel cladding, nitrogen impurities in fuel and fuel element core are 64%, 16%, 12%and 8%, respectively. The dominant part of 14C is induced by neutron reactions with coolant. The primary study results may offer some reference values for 14C releasing in environment control from NPP.
Thermal-Hydraulic Study of Wire-Wrapped Blanket Assembly of CEFR
Authors: Liu Yang , Yu Hong , Liu Yizhe
Published: February 2014

The blanket assemblies of China Experimental Fast Reactor (CEFR) were specially designed for the purpose of fuel breeding. There are two parts of the blanket assembly that have significant influence on its thermal-hydraulic feature: the labyrinth throttle set and the wire-wrapped fuel bundle. In this paper, the flow and temperature distribution of sodium coolant in fuel bundle, as well as the pressure drop in the labyrinth throttle set were investigated by solving the three-dimensional conservation equations of mass, momentum and energy, for a wide range of Reynolds number. Different turbulent models: k-ε, SST and Reynolds stress model (RSM) were used and compared with experimental correlations in the study of fuel bundle. The cross flow strength in three typical sub-channels and the influence of steer rods were also investigated. It has been found that the SST model fits best with the experimental correlation and although dampened the cross flow in edge channel, the steel rods installed in the hexagonal duct could enhance the heat transfer by “pushing” the sodium coolant to inner channel of the fuel bundle. Moreover, the relationship between the slot width of the labyrinth throttle set and the pressure drop under different coolant mass flow rate was analyzed, and found out the slot width that satisfied the design requirement of the blanket assembly.
Optimization Design for Air Heat Exchanger of Large Fast Reactor Based on Genetic Algorithm
Authors: Yu Huajin , Zhu Lina , Zhang Zhenxing , Liao Ziyu
Published: February 2014

The passive design for decay heat removal system of future fast reactor will put forward higher requirement for air heat exchanger (AHX), which is directly relevant to the structure and anti-seismic design of stack. Under considering the heat exchanger ability and the structure compactness comprehensively, a strategy for the optimization design of AHX based on genetic algorithm was developed in this paper. The air resistance in shell side of vertical fin tube AHX was chosen as the objective function, and the effect of design parameters including fin pitch, number of tube rows, tube pitch and tube length on the air resistance was discussed. The results of the study show that the method for the optimization design of AHX based on genetic algorithm can effectively optimize the structure of AHX and improve the resistance characteristic of the shell side evidently, which leads to design the fast reactor plant, stack structure and seismic resistance simply.
Research on the Measurement of Deep Subcriticality Based on Pulsed Neutron Source Method
Authors: Pan Shibiao
Published: February 2014

During the operating of nuclear reactor, the applicable condition of point reactor neutron kinetics is changed by control rods bulk insertion and reactor lattice deviating from the critical state, The measurement of reactivity, with the increase of subcriticality, shows that the results impact on the kinetic distortion effect, along with prompt neutron flux strongly deteriorated. According to the diffusion theory about prompt neutron and delayed neutron, the theoretical analysis and application of point reactor neutron kinetics have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that prompt neutron distributions are strongly affected by kinetic distortion. With the self-developed Pulsed Neutron Source measurement system, subcriticality measurement in different configuration of control rods in the zero power reactor of Nuclear Power Institute of China was carried out. In this paper the reactivity of the deep subcritical system are obtained by experiment of Pulsed Neutron Source method, meanwhile, the bias between experiment results of the areas-ratio method and the characteristic decay constant method are analyzed by comparing the condition of the experiment and the theory model, the main factors inducing the bias are found, which supplies helpful reference for other similar design and experiment.
Seismic Analysis of Lead-Bismuth-Cooled Fast Reactor Vessel
Authors: Ma Juan , Huang Mei
Published: February 2014

This paper is aimed at analyzing the structural responses of a Generation IV heavy-liquid-metal-cooled reactor (lead-bismuth-cooled fast reactor) vessel in the event of earthquake. For a seismic design, the seismic time history response analyses are carried out for both a top support type and a bottom support type. It is found that the bottom support type exhibits the better performance. There is the gap between the reactor vessel and guard vessel being filled with [1] argon. It is indispensable to add bellows at vessel upper end for a bottom support type to achieve the connection and seal between the reactor vessel and guard vessel. This paper is the first attempt to evaluate the effects on structural seismic performance by adopting the equivalent springs to simulate this elastic connection.

For a seismic investigation, there is a key issue that should be focused on, namely: the fluid-structure interaction due to seismic loading. The vessel is filled with a high-density fluid and might lead to severe hydrodynamic pressure significantly in the occurrence of earthquake. It will impair the structural integrity of reactor vessel. In order to study the structure effects of reactor vessel under seismic loading, an appropriate 3-D finite element analysis model has to be set up and the FEM code ANSYS has been implemented.
Study on Microwave Curing of Radioactive Waste
Authors: Xu Yanze , Ma Yan , Wei Yanni , Lu Chenlin
Published: February 2014

Processing of nuclear waste has become an important issue nowadays. A microwave waste disposal method for solid waste mainly containing Ni, Fe element and their oxide produced in the primary loop is investigated in this article. The powder waste, which is treated as absorber of microwave and energy source for curing, was doped in the curing agent that using SiO2 as a main component. In the experiment, using a household microwave oven (Microwave-frequency: 2.45GHz, Output power: 700W), thermal effects and non-thermal effects make samples heated to a high temperature quickly and then curing. In order to evaluate the effect parameters including different doping ingredient, heating period, pressure were changed. A pressure is applied to make a better curing effect. After curing, the measurement of the density of the cured body was made. The microstructure of the samples was observed with an optical microscope, and the cured body composition was analyzed by XRD and SEM. The results show that samples are sintered quickly with less than 15 minutes and the cured body has a good performance. It’s therefor that microwave technology can be an effective way to dispose microwave absorbing substances such as Fe, Ni-containing powder waste.
Research on Simultaneous Faults Diagnosis for Nuclear Power Plant Based on Dezert-Smarandache Theory
Authors: Wu Xiao-tian , Liu Yong-kuo , Liu Zhen , Xie Fei
Published: February 2014

Since the nuclear power plant is complex in structure and high-requirement in reliability, it is meaningful for the safety to build an intelligent faults diagnosis system. In recent years the Dezert-Smarandache Theory (DSmT) was put forward to relax the constraint of the exclusivity and provides a new way to deal with simultaneous and multiple faults. In this paper, a fusion model has been built and divided into three levels. The original data of the sensor has been decrease in data level. And then the BP neural network have been used for primary diagnosis and import the result to decision level as the generalized basic belief assignment of DSmT. After that the DSmT fuse all the evidence and make a decision. Finally, the method was tested by the data of SGTR, LOCA and main pump accident. And the results indicate that it can diagnose not only single fault, but also simultaneous and multi-faults accurately, and has some application value.
Local Mesh Refinement for Viscid Flow in Combustion Code COM3D
Authors: Ren Ke , Kotchourko Alexei , Lelyakin Alexander
Published: February 2014

After the implementation of local mesh refinement (LMR) in solution of Euler equations and detonation simulation, the technique has been used to optimize the simulation of viscid flow. In this paper, LMR has been implemented in Navier-Stokes (NS) equations and LES turbulent model. In order to implement LMR in the simulation of viscid flow accurately and efficiently, hybrid interpolation and coarse flux limiter are established. At the same time, implementation of LMR maintains the temporal refinement in solution of NS equations. In practical application, 2D advection diffusion is used to show the advantages of LMR and verify the implementation.
Numerical Analysis of Containment Response Under Accident Conditions
Authors: Yu Shengzhi , Wang Jianjun
Published: February 2014

On the basis of passive containment cooling system concept, the one-dimensional codes are developed for the analysis of containment thermal-hydraulic characteristics under accident conditions, which can be applied to deal with large dry concrete containment in nuclear power plant. In order to build up the hypothesis flowing path, the containment space is divided into a rising channel and downward ring during the geometric modeling process. In this paper, the physical models, the identification of solving methods and the verification of the codes are introduced. It is assumed that the control volumes take the adiabatic condition in addition to heat exchanger and breaks, and there is no exchange of mass, momentum and energy between each volume in the rising channel and corresponding volume in downward ring. In the analysis, we also assume that the heat in the containment can only be transferred through natural circulation by passive containment cooling system. Furthermore, the break is supposed in the center of bottom of the containment. In this paper, the responses of the containment are predicted with the codes under large LOCA scenario. Under the same conditions, the characteristics of the natural circulation are also analyzed through the codes for the passive containment cooling system. The results can provide some references for the design of the passive containment cooling system.
Study on Operating Criteria of the Supersonic Steam Injector
Authors: Satoh Kyosuke , Abe Yutaka , Kaneko Akiko
Published: February 2014

Supersonic steam injector (SI) is a passive jet pump with a venturi structure which operates by condensation of the supersonic steam flow in a subcooled water jet. There are advantages that a SI which operates without electric power has high heat-transfer performance and discharges water at high pressure. Therefore the SI is expected to be applied to the safety system which is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant.

According to the previous studies about a SI, there are two conditions about the definition of a SI, one is formation a water jet in the mixing nozzle, the other is stopping drainage from the drain. The 1-dimension analysis model, which the steam condensed completely, at the mixing nozzle, is developed by Narabayashi et al. (1996). Abe et al. (2011) quantified the interfacial behavior of water jet with high resolution observation in the mixing nozzle and evaluated correlation between the interfacial behavior and heat and momentum transfer characteristics. However the operating limit of a SI which is taken account of flow structure at the diffuser is not cleared.

The objective of the present study is to reveal the effects of two-phase flow behavior in the diffuser on the operating limits of the SI. In the present study, static pressure distributions along flow direction were measured in the SI with change about the back pressure. The discharge pressure was measured and compared with the value of discharge pressure predicted by 1-dimention analysis model. On the other hand, the discharged flow in the diffuser was observed with a high speed camera.

According to the measurement of pressure, a significant pressure rise in the diffuser was measured as the back pressure increased. In addition, the value of the discharged pressure just before the SI becomes inoperative by the back pressure rise was defined as the maximum discharge pressure. It is suggested that this maximum discharge pressure is one of the indicators which predicts the operating limit of the SI. From visualization result of the diffuser, it was confirmed that there is propagation of white color. The void fraction in the discharge flow was estimated at about 2% from the velocity of the propagation.

From the experimental results, effect of two-phase flow behavior of discharged water jet on the operating limits of a SI is discussed.
A Review on Application of Technologies for Measuring Hydrogen Concentration in Containment
Authors: Luo Sha , Wang Shaobo , Qin Liang , Pang Feng
Published: February 2014

Hydrogen concentration measuring technologies in containment can play an important role in preventing radioactive materials leakage and ensuring the safety of people and property in nuclear accident. However, as a confined space, the environmental condition in the containment during a severe accident is very complex and harsh, with high temperature, high pressure, high humidity and high-level radioactive. Thus, it is an issue worth exploring how to ensure the availability, safety and measurement accuracy of the hydrogen measuring devices in such special conditions. This article reviews the hydrogen measuring technology and related products currently used in nuclear power plant in the world. Because of the particularity of the severe accident conditions, some measuring technology and related products need further improvement and perfection. After the Fukushima nuclear accident, the technology for measuring hydrogen concentration in containment has become more compelling. The CH-15 type hydrogen concentration measuring device, which was independent developed by the 718th Research Institute, was briefly introduced in the end of this article. With deepening understanding of nuclear accident by nuclear power industry, the advanced hydrogen measuring devices will be widely used in nuclear power plant in the world, which will be helpful to improve the safety level of nuclear power plant.
Self-Stability Analysis and Nonlinear Power-Level Control for Pressurized Water Reactors
Authors: Dong Zhe
Published: February 2014

Power-level regulation is a key technique that provides both economic and stable operation for any nuclear reactors. Since the pressurized water reactors (PWR) are not only the most widely utilized nuclear reactors but also complex nonlinear systems, efficient nonlinear power-level control strategy is meaningful to not only the development of nuclear power technique but also the renaissance of nuclear energy. Further, since self-stability is the most crucial dynamic feature of nuclear reactors, it is necessary to give self-stability analysis which gives the means to strengthen the closed-loop stability. In this paper, a sufficient condition of globally asymptotic self-stability for PWR is firstly given. Based on this self-stability analysis, a nonlinear state-feedback power-level control law with L2 disturbance attenuation performance is then proposed. This new controller is applied to the power-level regulation of a nuclear heating reactor (NHR), and simulation results show its feasibility and high performance.
Study on Flushing Phenomena by Microwave Heating
Authors: Yamaki Tatsunori , Kaneko Akiko , Abe Yutaka , Segawa Tomoomi , Kawaguchi Koichi , Yamada Yoshikazu , Suzuki Masahiro , Fujii Kan-ichi
Published: February 2014

Recently, the number of nuclear power plants has been increased in many countries. In contrast, uranium fuels used in nuclear power plants are exhaustible resources. Therefore, it is required to exploit uranium resources effectively, and reprocessing of spent fuel is indispensable. To use recovered uranium and plutonium as raw material of nuclear fuel, reprocessing solution (uranium and plutonium mixed nitrate solution) of the spent nuclear fuel is converted to uranium and plutonium mixed oxide (MOX) powder. Microwave heating direct denitration method (MH method) is one of such methods to convert nitrate solution to MOX powder. The cylindrical denitration vessel can be expected to realize high-speed and high-capacity processing against traditional shallow vessel. However, flushing and overflow phenomena of solution have been confirmed in cylindrical vessel. Thus, the safety and the optimization of the vessel shape during microwave heating. In the present study, the purpose of this paper is to clarify generation conditions and generation mechanism of flushing phenomena that is not fully understood. In experiment, flushing phenomena was observed and the liquid temperature was measured using microwave heating device. The main parameters are the vessel diameter, initial water level and the magnetron power.

There was tendency of flushing in the case of short vessel diameter and high initial water level when magnetron power was constant. There was also tendency of flushing in the case of large magnetron power when vessel diameter and initial water level was constant. From visualization, it was clarified that generation of singular bubble triggers flushing. If flushing occurs, the liquid is blown up at a burst, and the vessel become almost empty. From temperature measurement results, it was clarified that the liquid had over 10 °C superheat just before flushing. Therefore, it was suggested that the liquid superheat affected flushing significantly. Generation conditions of flushing are different with the vessel diameter, initial water level and the magnetron power because it is considered that these characteristics have influence on the liquid superheat.
The Development and Application of PWR Severe Accident Simulation Model
Authors: Chen Yanfang , Xie Zhengquan , Lin Xusheng , Shan Fuchang , Wei Wei , Guo Fude
Published: February 2014

The severe accident simulation codes that developed by the engineers in RINPO are introduced in this chapter. The results of the severe accident caused by large LOCA plus losing safety core injection are presented. Comparison with the results of SCDAP/RELAP5/MOD3.2 of the same accident and the same nuclear power plant type has been made. From the comparison and the analysis we can make the conclusion that the trend-lines are correct and the mathematical models are reasonable in this simulation code.
Experimental Study on Sodium Column Fire of Sodium-Cooled Fast Reactor
Authors: Peng Kang-wei , Zhang Zhi-gang , Guo Ming , Wang Chao , Sun Shu-bin
Published: February 2014

In the operation of sodium-cooled fast reactor, accident caused by the leakage and combustion of liquid sodium is common, and it is a pivotal and difficult problem in research, construction and operation of sodium-cooled fast reactor internationally. In actual operation of sodium-cooled fast reactor, liquid sodium in sodium fire accident is difficult to form fog but mainly in columnar flow form due to low operation pressure and thermal insulation material wrapping the pipe, so experimental research about columnar fire is of much more practical significance.

This paper focuses on combustion property on sodium column fire in the sodium-cooled fast reactor. Liquid sodium with high temperature will be poured into the combustion room via pouring high pressure nitrogen into sodium storage tank when solid sodium has been heated to enactment temperature; liquid sodium with high temperature will burn at combustion room with air forming sodium column fire. Initial temperature of sodium jet is about 200°C in experiment and spurt pressure is 0.2MPa, spurt flow is about 0.4–0.6m3 each hour. Temperature fields in combustion tank space and the catch plate in the bottom are measured through dozens of thermocouple distributed in combustion tank. No atomization phenomenon would exist in the pure sodium columnar fire in the space, the atomized fire triggered by splashed sodium from sodium stream striking the admittance plat and generate more than the high temperature of 900°C.
The Investigation on Interactive Products for Propaganda of Nuclear Safety in Public
Authors: Hou Fei
Published: February 2014

Depend on the development of technology, the interaction products have been gradually applied into public awareness in the part of Nuclear Safety. Therefore how to put the information about Nuclear Safety into entertainment produces become a crucial research direction in related industry. The study is to investigate current interaction products based on collection of related information in order to find valuable results for effective propaganda on Nuclear Safety.

After the nuclear accident happened in Japan, the popular distrust of Nuclear Energy is spreading all over the world. Due to the kind of negative voices from public in China, several Nuclear constructions have been suspended. Consequently public awareness on Nuclear Safety is supposed to gain more attention for developing Nuclear energy.

According to the investigation on current products on Nuclear Safety, the most of exhibited products are unattractive. How to integrate knowledge of Nuclear Safety into interactive technology becomes a key question. The interaction products need to be improved by designing, developing and manufacture. As a result of complex production process, one interaction product needs interdisciplinary cooperation, so that the future products could become attractively, entertainingly, effectively, pertinently from the conclusions of this research.

Generally, this study is aim to collect the information of interaction products on Nuclear Safety in the whole world, then investigate and analyze current status using appropriate research method. Finally valuable results might be presented for improving interaction products on Nuclear Safety from the whole production process.
Three-Dimension Thermal Hydraulic Analysis of Stationary Target Plate for Highly Intensified Neutron Generator (HINEG)
Authors: Li Shuzhou , Chen Hongli , Zhou Tao , Song Fengquan , Wang Wen
Published: February 2014

Highly intensified neutron generator (HINEG) is a D-T neutron generator tritium target system; it can be used in researching fusion energy and advanced fission energy. The heat flux at the target plate is extraordinarily high, cooling the target plate effectively, limiting the working temperature, are the keys to keep the normal working of HINEG. This paper has simulated the 3D values of some different kind of target plate systems, using the computational fluid dynamics software and static analysis software, presented the stationary analysis results of temperature and fluid fields, as well as stress field. The flow and temperature distributions provide important data for advanced design and performance evaluation of HINEG.
Experiment Study on Collision of Graphite Dowel-Brick Structures
Authors: Jin Lie , Sun Libin , Wang Hongtao , Wang Haitao , Wu Xinxin , Ma Shaopeng
Published: February 2014

Graphite bricks have important applications in high temperature gas-cooled reactors (HTGRs), the core of HTGR is a pile-up of graphite bricks. So the vibrations and collisions between graphite bricks caused by external excitation have an important influence on structural stability of the core. The locations of bricks are fixed by various kinds of keys and dowels. The collision experiment, with tracks and small railcars as experimental devices and measurement system using optical method, was aimed at studying non-central collisions between two bricks. The passive one of the two bricks was equipped with a key or a dowel. Experiment’s results revealed how the coefficient of restitution and the contact time would change within the range of velocities of the active specimen. It was showed that the contact time would increase with the rise of initial velocity while the coefficient of restitution would rise up firstly and then decrease later in the same process. Besides, qualitative influence caused by different sizes of keys and dowels was briefly discussed, and material properties of graphite was not the dominate factor in the collision of dowel-brick structures experiment, while the velocity of active specimen just before collision and the fact that the collision is non-central have more significant effects on the collision results.
A Novel Method for Controller Parameters Optimization of Steam Generator Level Control
Authors: Kong Xiangsong , Wang Xu , Ma Zhicai
Published: February 2014

Control performance is a key to steam generator level control system. To improve the system’s performance, the parameters of the control system should be optimized. Traditional parameters tuning methods are experience-based, cumbersome and time-consuming. A novel optimization method based on the Simultaneous Perturbation Stochastic Approximation (SPSA) is proposed to maximize the system’s control performance. Rather than traditional controller parameter tuning method, this method optimizes the control system by directly using measurements of control performance. An iteration termination control strategy is developed and integrated for the parameters optimization. The effectiveness of this method has been demonstrated through simulation experiments.
Research on Bubble Growth and Frequency at Higher System Pressure in Narrow Rectangular Channel
Authors: Hu Jian , Gao Puzhen , Zheng Qiang
Published: February 2014

In present work, the characteristics of bubble growth and frequency were visually observed in a narrow rectangular channel using high-speed digital camera. The experiment was done over the following range of conditions: pressure, 0.55MPa; mass flux, 300–500kg/(m2·s); heat flux, 86.4–225.7 kW/m2; and inlet subcooling, 25.5–45.5K. The system pressure has a significant effect on the bubble growth. Experimental results show that the bubble diameters are just about 0.15mm under higher system pressure and the period during which bubbles attaching to the nucleate site is very short, or even unnoticeable. The bubbles keep growing when slide along the heating wall rather than lifting off the surface, and the bubbles rarely collapse under the working conditions. When the mass flux is high, the bubble diameter increases with increasing the heat flux, but when the mass flux is low, the variation trend of bubble diameters expresses no obvious law. The effect of thermal parameters on bubble frequency is also significant. When the mass flux is low, the bubble frequency decreases with increasing the heat flux or inlet subcooling, however when the mass flux is high, the bubble frequency increases first and then decreases. Generally, the bubble frequency increases with increasing the mass flux.
Application of Integrated Safety Assessment Methodology to SBO Sequences
Authors: Queral C. , Mena-Rosell L. , Jiménez Varas G. , Sánchez-Perea M. , Hortal J. , Meléndez E. , Gómez-Magán J. , Gil J. , Fernández I.
Published: February 2014

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest as a function of the operator actuations times (recovery of AC). Given a particular safety limit or damage limit, several data of every sequence are necessary in order to obtain the exceedance frequency of that limit. In this application these data are obtained from the results of the simulations performed with MAAP code transients inside each damage domain and the time-density probability distributions of the manual actions. Several damage limits have been taken into account within the analysis: local cladding damage (PCT>1477 K); local fuel melting (T>2499 K); fuel relocation in lower plenum and vessel failure. Therefore, to every one of these damage variables corresponds a different damage domain. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.