bluuebunny/crossref_metadata_2025 huggingface dataset:
Nuclear Safety Culture Construction in Nuclear Power Design Enterprise on the Perspective of Knowledge Workers in China
Authors: Li Qiu , Wu Jun
Published: February 2014

Design plays a leading role in the whole life cycle of nuclear power production, including site selection, design, manufacture, construction, operation and decommission. The quality of the design products directly affects the intrinsic safety of the nuclear power plant. The quality of final products in design process depends on a number of factors, including not only technology capability, but also the improvement of design quality assurance system, the capability and responsibility of the designers to nuclear safety, the thinking mode and working habit of the designers, and the extent of complete implementation of the review systems at all levels. These factors consequently affect the level of safety of the nuclear power plant.

Nuclear power design industry is typical knowledge-intensive business, in which knowledge workers consist of essential assets to the enterprise. The article analyzes the asset structure and staff structure of nuclear power design industries in contrast with other industries, and discusses the contribution of knowledge workers to the development of the enterprise. This paper also tries to document the characteristics of knowledge workers in nuclear power enterprises. They are characteristic of superior income and welfare, high level education and ability, high specialization, and tacit knowledge.

Based on these analyses, the article addresses four major principles in nuclear safety culture construction for nuclear power design enterprises as follows: 1. adopt motivation factors as major incentives, 2. emphasize self management, 3. management by objectives (MBO), 4. team work.
The Research of the Fluid-Solid Interaction of the Passive Containment Cooling Tank and Shield Building Structure With Seismic Load in AP1000
Authors: Dang Junjie , Lu Daogang , Ma Wenhui , Liu Yu , Hong Yang
Published: February 2014

The passive containment cooling system is the ultimate heat trap and the key to ensure the safe operation of the AP1000 nuclear power plant. It is very important to keep the reliability of the passive containment cooling system (PCS) in the normal conditions and also in the abnormal conditions, especially in earthquakes. Analysis on the fluid-structure interacted characteristics of the passive containment cooling water storage tanks and shield building structure in earthquake is crucial to the assessment of the effectiveness of the PCS. According to the passive containment cooling water storage tanks and shield building structure prototype size, an experiment model was established to analyze the fluid-structure interacted characteristics of the passive containment cooling water tanks and shield building by the numerical method.
Improving the Accuracy of Gamma Spectrum Analysis by Total Variation Based Adaptive Smoothing
Authors: Sun Sida , Fang Sheng , Zhang Liguo , Li Hong , Tong Jiejuan , Qu Jingyuan
Published: February 2014

Gamma spectrum analysis is an important part of gamma-ray spectroscopy which has been widely used in nuclear engineering, environmental science and astrophysics. As the first step of gamma spectrum analysis, smoothing procedure is critical, since it determines the accuracy of afterwards procedures. Ideally, the smoothing procedure should reduce the statistical fluctuation while preserving characteristic peak information simultaneously. However, current widely-used linear smoothing methods are intrinsically non-adaptive and tend to remove weak peak, which may lose characteristic peak of important radionuclides. To solve the problem, an adaptive smoothing method was proposed to improve the accuracy of gamma spectrum analysis in this study. The proposed method assumes that gamma spectrum is a sparse signal that has meaningful peaks only at limited positions. Based on this assumption, the smoothing procedure is formulated as a nonlinear total variation based optimization problem. Solving this problem promotes the sparsity of gamma spectrum, and therefore reduces meaningless fluctuation, so that the spectrum is adaptively smoothed. The proposed method was applied to gamma spectrum obtained by a Monte Carlo experiment that simulated the ORTEC GEM3070 detector, and compared with traditional linear method. The results demonstrate that the proposed method can effectively reduce the statistical fluctuation of measured gamma spectrum while preserving weak peak much better than standard linear methods. With the proposed method, the accuracy of peak identification and peak calculation is significantly improved.
Low-Level Liquid Waste Treatment System Technical Design in China
Authors: Zhao Juan
Published: February 2014

Radioactive wastes are produced within the nuclear fuel cycle operations (uranium conversion and enrichment, fuel fabrication and spent fuel reprocessing). Evaporation is a proven method for the treatment of liquid radioactive waste providing both good decontamination and high concentration. Two technical designs of nuclear facilities for low-level liquid radioactive waste treatment are presented in the paper and the evaluation of both methods, as well. One method is two-stage evaporation, widely used in the People’s Republic of China’s nuclear facilities; another is two evaporator units and subsequently ion exchange, which is based on the experience gained from TIANWAN nuclear power plant. Primary evaporation and ion exchange ensure the treated waste water discharged to environment by controlling the condensate radioactivity, and secondary evaporation is to control concentrates in a limited salt concentration.
The Design and Modification to Radiological Controlled Area Gateway of Fuqing NPP
Authors: Yang Wei , Huang Hong , Hu Ling
Published: February 2014

In PWR NPPs, the RCA (Radiological Controlled Area) gateway’s design is a very important factor to the effectiveness of RCA boundary control and on-site radiation protection management. This issue summarized the practice of RCA gateway’s design & modification of unit 1 & 2 in Fuqing NPP, with which a presentation on constitution, function and basic principles of design and layout, what would be a reference for the coming advanced PWR reactors.
Design of Moderators and Reflectors for Chinese Spallation Neutron Source
Authors: Du Wenting , Hu Chunming , Yu Chaoju , Tong Jianfei
Published: February 2014

Chinese Spallation Neutron Source (CSNS) is a significant research facility for studies of structure and dynamics of materials, which can play an important role in research and education. Moderators and reflectors system is one of the core components in CSNS. There are three moderators totally, namely, the decoupled and poisoned hydrogen moderator, the coupled hydrogen moderator and the decoupled water moderator. Moderators are inserted into an aluminum alloy vessel which contains beryllium and stainless steel reflectors. The moderators and reflectors are designed into a whole plug, MR Plug, so as to move the plug into or out of the core vessel of CSNS easily. The structure design of moderators and reflectors system is just finished now, including moderators, reflectors and their container. Neutronics requirements and strength criterions are satisfied, manufacture of system is feasible. Structure strength analyses of some parts are also presented.
An Implementation of DFM for Reliability Modeling and Analyzing of AP1000 FWCS
Authors: Yin Jiubo , Cao Jianyuan , Wang Xu
Published: February 2014

Both in the process of designing instrumentation and control (I&C) systems of nuclear power plants and replacement of aging and obsolete ones, digital technology are more and more popular. However, traditional static methodology for probabilistic risk assessment (PRA) may not yield satisfactory results when applied to digital systems. As a result, some dynamic methodologies are employed in the analysis for I&C systems of nuclear power plants. The dynamic flowgraph methodology (DFM) is one of the favorable methodologies which brings out most positive features and least negative features. The I&C system of AP1000 is a distributed digital I&C system, but its PRAs are based on the event tree (ET)/fault tree (FT) approach. The results are admitted to be enlarged to some extent. Also, under some circumstances, ET/FT approach cannot account for the coupling between triggered or stochastic logical events. Regarding these disadvantages, DFM is introduced in modeling of Feed-water Control System (FWCS) of AP1000. A basic system of FWCS is constructed, which contains the main components and control logic of FWCS. Furthermore, failure mode and effect analysis (FMEA) of the FWCS system is carried out in the way of analyzing each component individually. After these efforts, a DFM model of the basic system is built utilizing the DFM language. After the DFM model has been produced, the application of deductive analysis shows different initial conditions or events that can cause the FWCS system to fail in different modes.
Application of RSE-M2010 on In-Service Inspection of Taishan EPR Project
Authors: Zhang Xueliang , Shao Chunbing , Tang Ximing , Yang Cheng , Feng Huixing
Published: February 2014

The latest edition of French In-service Inspection Rules RSE-M2010, incorporating the up-to-date upstream French regulations, orders and requirements for pressure equipments, and taking into account both of the radioactive risk and industrial risk in nuclear power plant (NPP), has been adopted as the applicable rule for in-service inspection (ISI) of EPR units. In RSE-M2010, the previously used benchmark for classification Safety Class has been replaced by the Nuclear Pressure Equipments Class (ESPN Class), and the category of pressure equipments has been introduced to monitor the industry risks of NPP pressure equipments, making it much more precise and convenient to define the scope of equipments which subjected to ISI and corresponding ISI requirements on frequency and methods. This paper described the main differences of the ISI requirements in RSE-M2010 and previous edition of RSE-M, also introduced practices of applying RSE-M2010 when preparing the ISI program of Taishan EPR units. Based on the application practice of RSE-M2010 on Taishan EPR project, some proposals for future improvement of this code are presented. Preliminary thinking for future implementation of EPR ISI activities has also been described.
Physics Analysis of the Accelerator Driven Subcritical Reactor Core
Authors: Wang Feng , Qin Xue , An Zilong , Cui Bo
Published: February 2014

The reactor core of an accelerator driven sub-critical system has been physically analyzed by the MCNP code. Neutron flux density of different area within the reactor has been calculated, and the influence on its distribution has also been analyzed. Results show that there exists higher fast neutron flux variation at different element layer in fast region, and relatively lower thermal neutron flux variation at different element layer in thermal region. The calculated neutron flux meets the general design requirements in the reflector and shielding layer. Neutron multiplication factor is remarkable in the fast neutron spectrum area, and it realizes the energy amplification in the thermal spectrum area. The statistical particle number of code can influence the accuracy of the calculation and variation of the core design parameters can change the neutron flux distribution in the reactor core.
A Study of Thermal Aging Effects on Microstructures of Dissimilar Metal Weldment
Authors: Choi Kyoung Joon , Kim Jong Jin , Kim Ji Hyun
Published: February 2014

Characterizations of microstructure and mechanical property were conducted to investigate the effect of long-term thermal aging on the microstructural evolution in the fusion boundary region between weld metal and low alloy steel in dissimilar metal welds. A representative dissimilar weld mock-up made of Alloy 690-Alloy 152-A533 Gr. B was fabricated and aged at 450°C for 5,500 hours. It was observed that the region near the fusion boundary was generally divided into several regions including dilution zone in the Ni-base alloy weld metal, fusion boundary, and heat-affected zone in the low alloy steel. A steep gradient was shown in the chemical composition profile across the interface between A533 Gr. B and Alloy 152. The precipitation of carbides was also observed along and near the fusion boundary of aged dissimilar metal joints. It was also found that the precipitation of Cr carbides was enhanced by the thermal aging near the fusion boundary.
The Fracture Behavior of Cylinders With Axial Surface Cracks Under Strong Through-Wall Bending Stresses
Authors: Li Yuebing , Gao Zengliang , Lei Yuebao
Published: February 2014

Reactor pressure vessel (RPV) may subject to strong through-wall bending stresses under some situations, such as pressurized thermal shock (PTS) transients. The fracture behaviors of cylinders with internal axial surface cracks under PTS conditions are investigated in this paper. A typical PTS transient is selected to simulate the bending stress. The critical crack sizes including crack depth and crack surface length for a given cylinder is analyzed for the assumed PTS transient using the critical crack depth diagram method. Crack initiation and growth at both the deepest and surface points are considered and the effect of the crack growth at surface point on the growth along the wall-thickness is investigated. The results show that the crack initiation and growth at the surface point should be considered and the critical crack depth for the assumed transient may depend on the fracture behaviors at both the deepest and surface points.
Two-Way Coupling Simulation of Heat Transfer in ADS Windowless Spallation Target
Authors: Liu Jie , Gao Lei , Chen Huan , Wang Zeng-hui , Lu Wen-qiang
Published: February 2014

The steady-state phase distribution and the structural parameters have been taken as the input for the nuclear physics calculation in the ADS windowless spallation target. The distribution of the extreme large power density of the heat load is imported back as the source term in the energy equation. Then temperature distribution is obtained based on the flow process and heat transfer. The preliminary results show that the temperature distribution reaches the steady-state and its shape is like the broken wings of the butterfly. This is very important for the further design and optimization of the ADS windowless spallation target. So the two-way coupling simulation of the heat transfer process is successfully performed between the computational fluid dynamics and the nuclear physics simulation.
Startup of In-Hospital Neutron Irradiator
Authors: Li YiGuo , Xia Pu , Wu XiaoBo , Zou ShuYun , Peng Dan , Lu Jin , Hong JingYan , Zhang ZiZhu , Liu Tong , Zhou YongMao
Published: February 2014

In-hospital Neutron Irradiator (IHNI) was specially designed for Boron Neutron Capture Therapy (BNCT), the rated power of IHNI is 30kW, corresponding to the neutron flux density 1×1012n·cm−2·s−1 in reactor core. IHNI is an undermoderated reactor of pool-tank type, and UO2 with enrichment of 12.5% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection. On the both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the Prompt Gamma Neutron Activation Analysis (PGNAA). The decay constants and shares of six group of ordinary delayed neutron and nine group of photoneutron were obtained by WIMS code. Based on that, the relationship between the reactivity and the reactor period was calculated through the inhour equation. In this way, the excess reactivity and the reactivity worthies of the components (control rod, water, etc) in the core are obtained by periodic method during the startup of the reactor. The six test experiments were completed during startup, The test results show that the maximum continuous operation time at full power is 12h; the excess reactivity at cold clean state of the core is 4.2mk; The radiation levels at technical rooms are within the specified values at full power operation. When the positive reactivity with 4.2 mk is inserted into the reactor suddenly, the power will be increased to peak power, and then, it will turn to the normal value due to the negative temperature effect, this experiment shows the inherent safety of IHNI.
Device for Measuring Hydrogen Concentration in Containment During Severe Accident
Authors: Luo Sha , Wang Shaobo , Qin Liang , Pang Feng
Published: February 2014

The hydrogen explosion in Fukushima nuclear accident seriously challenged the safety of nuclear power plants in the world. Therefore, strengthening hydrogen concentration measurement in containment has become very important. To raise the nuclear safety level of in-service and under construction nuclear power plants in China, based on a document which was issued by the National Nuclear Safety Administration of China (NNSA, 2012), the function requirements, components, storage, and arrangement outlines of the hydrogen monitoring system after a severe accident was thoroughly analyzed in this article. Besides, two kinds of techniques for hydrogen concentration measurement in containment, respectively for direct measuring method and gas sampling measuring method, were also discussed and compared. Notably, based on the direct measuring method, the 718th Research Institute successfully developed the CH-15 type hydrogen concentration measuring device that suitable for used in severe accident. The hydrogen sensor, with independently intellectual property rights, is based on catalytic principles and installed inside the containment. It has many unique characteristics, such as wide measurement range, high measurement accuracy, and is capable of continuous measurement with multiple points. This device has simple structure and small size, with low energy consumption, so it is very suitable for being installed in the in-service and under construction nuclear power plants in China and abroad.
Design Analysis of Corium Immobilization and Cooling of the Reactor Vessel in Severe Accidents at NPP With Commercial Unit V-320
Authors: Stremedlovskyi Roman A.
Published: February 2014

Keeping the melt inside of the reactor vessel is the key strategy of the management of severe accidents for the exploited, as well as for the newly designed reactors with pressurized water. The concept has been substantiated for emergency cooling of VVER-440 reactor at Loviisa NPP in Finland [1]. However, for more powerful reactor, such as VVER-1000, up to now it was considered that this concept did not work.

This work analyzes the problem of the interaction of molten corium with the reactor vessel. The work is representing the results of studies that were carried out as part of research with a goal to validate design functions of the system of containment the melt inside the VVER vessel.

The VVER or WWER is a series of PWR designs originally developed in the Soviet Union, and nowadays in Russia. The reactor plant of VVER-1000 type of V-320 series is a reactor plant with a water-cooled and water-moderated power reactor for NPP. It has an electric power of 1000 MW and a thermal power of 3000 MW. The plant consists of four loops, each one with a horizontal SG and a MCP. In the work the design analysis of the accident with melting of the core of the reactor plant V-320 was conducted on codes RELAP/SCDAPSIM/MOD3.4 and MELCOR. The main processes that determine the state of the core melting in the reactor vessel were analyzed. The main results of the calculation of the molten corium interaction with the wall of the reactor vessel were represented. The results of the preliminary calculated estimation of the melt behavior and the technical ability to hold the core melt in the vessels of PWR reactors were represented.

Based on the carried out analysis and implementation of technological and design study, the recommendations were made with a purpose to improve the safety of NPP with a reactor plant V-320 using the systems of melt restraining and cooling of the reactor vessel.
Neutronics Design of a Small LBE Cooled Accelerator Driven System CIADS
Authors: Wang Xinzhe , Yu Hong , Zhang Jian , Wang Shixi , Hu Yun , Duan Tianying
Published: February 2014

As an important part of advanced fuel cycle R&D, conceptual study of accelerator driven system (ADS) in China started since 1995. In 2000, China Institute of Atomic Energy (CIAE), Institute of High Energy Physics (IHEP) and other institutes started a ten-year project aiming at ADS fundamental R&D on physics and related technologies, which is one item of “Key Project of Chinese National Program for Fundamental Research and Development (973 Program)” in energy domain. In order to get a better understanding of ADS neutronics characteristic, China Fast Reactor Research Center initiates a preliminary R&D program focused on neutronics design of a small lead-bismuth eutectic cooled ADS with fast spectrum. In this program, the reactor core of a 10MW thermal power ADS called CIADS (China Initiative ADS) with MOX fuel has been studied and designed. For generally concerning, CIADS can operate in either subcritical or critical mode. Different parameters, such as target size and position, position that transmutation assemblies are placed have been studied during the design work. Results show that a half size target and one zone loading can meet the needs for a small size ADS. Moreover, some important physical parameters of CIADS, such as keff, ks, power peak factor and neutron maximum flux density are evaluated. According to the R&D work, it’s appropriate to set the ks of CIADS at 0.96∼0.98.
Primary Shielding Design and Analysis for CLEAR-I
Authors: Yang Qi , Li Bin , Chen Chao , Wang Minghuang , Zeng Qin
Published: February 2014

The China Lead-Alloy Cooled Research Reactor (CLEAR-I) is critical/sub-critical dual-models natural circulation lead alloy cooled reactor. This study is to focus on the concern radiation shielding design and analysis for CLEAR-I. The modeling program MCAM and calculation system VisualBUS developed by FDS Team was used based on Monte Carlo method and other coupled methods. As indicated by the results, the dose rate in the reactor plant (outside the containing compartment above the reactor) was below 9 uSv/h during operation and less than 1 uSv/h during shutdown, meeting with the requirements of shielding.
General Design Study on CSR1000 Structure
Authors: Zhang Hong-liang , Fan Heng , Luo Ying , Li Xiang , Liu Xiao , Zhou Yu
Published: February 2014

Based on China Supercritical-Water-Cooled Reactor 1000 (CSR1000) research project, this paper introduces the general design research progress and the structure design scheme of the two-pass reactor, of which the key technologies are studied or proposed, including reactor structure material, sealing type, flow distribution, thermal stress analysis and flow-induced vibration response analysis. Basic research methods and solutions are provided, which are meaningful for the engineering practice.
Simulation Research on Combustion of Liquid Sodium Droplet
Authors: Wang Chao , Zhang Zhi-gang , Peng Kang-wei , Guo Ming , Sun Shu-bin
Published: February 2014

Since sodium leak and sodium fire belongs to incidental accident in fast reactor. This paper made the simulation research on sodium droplet combustion, which was the most fundamental one of the analysis on sodium spray fire. Referencing to the sodium droplet combustion experiment, this paper made improvement in the traditional model of the sodium droplet and researched on the initial temperature and oxygen fraction on oxidation behavior of sodium droplet.

This paper used the new spherical model with cones on the sodium droplet surface and improved the spherical model used in the traditional calculation of sodium droplet. This made the simulation was closer with the actual combustion situation.

According to the characteristics of the reaction between the sodium and oxygen, this paper modeled sodium droplet combustion process, which was divided into three stages, namely the surface oxidation stage, pre-combustion stage and combustion stage, set up the mathematical model, compiled the calculation program and got the regulation of the sodium droplet temperature changed with time under different initial temperatures (200–500 °C) and different oxygen fractions (4–20%). This program was validated with the sodium droplet combustion experiment. Through using the proportion γ of the sodium mass participated in the oxidation reaction to vaporizing reaction in the unit time, under the different conditions of sodium droplet combustion, this paper used different γ in different reaction stage, got the simulation results, which were consistent with the experimental data and analyzed the changing regularity of γ.
Study on the Nonradioactive Environmental Impact Assessments for Chinese Nuclear Power Plants
Authors: Wei Xinyu , Ma Hongbin , Xiong Xiaowei , Yang Duanjie , Shang Zhaorong
Published: February 2014

In China, the radioactive environmental impact assessment (EIA) for nuclear power plants (NPPs) had been carried out in detail due to the specific characteristic of radiation and the social universal attention. However, the nonradioactive environment impact assessment for NPP doesn’t get enough attention. This should be improved, since the operation of NPPs could cause some serious nonradioactive environment impacts.

Based on the investigation of EIA for American NPPs, the following suggestions were put forward for Chinese EIA: (1) the laws and regulations in China need to be revised for the EIA of NPPs; (2) the statistics of chemicals or biocides used in the operation of NPPs and their discharged concentration need to be tested; (3) the impact assessments of nonradioactive gas emissions as well as hazardous wastes need to be carried out to meet the national requirements; and (4) the monitoring of nonradioactive waste also needs to be carried out for NPPs to obtain the basic data.
Investigation of Transition Boundary and Nature of Pulsating Flow in the Trickle Bed Reactors by Electrical Resistance Tomography
Authors: Eda Takeshi , Sapkota Achyut , Haruta Jun , Nishio Masayuki , Takei Masahiro
Published: February 2014

A fixed bed reactor that operates in gas-liquid co-current down flow is called Trickle Bed Reactor (TBR). It is widely used in chemical engineering. And, recently used in purification of radioactive contamination from contaminated water generated in the Fukushima Daiichi nuclear power plant. There are several flow conditions that occur in the TBR due to gas and liquid flow rate. Since mass and heat transfer rate and particles wetting depend on flow condition, it is necessary to establish the visualization techniques to understand flow condition, transition boundary and properties of gas liquid flow in TBR. In this study, authors employed the lab-scale TBR, made of 100mm inner diameter acrylic column, packed with particles of two sizes (3, 5 mm) that are used in the actual reactor. Water and air were injected from the top of the column and cross-sectional liquid distribution was captured at the bottom of the column by electrical resistance tomography (ERT). ERT is a tomographic technique that provides the cross-sectional conductivity distribution at the rate of about 50 frames per second by injecting current and measuring voltages between the 16 electrodes that are attached around the column. By analyzing the spatial and temporal characteristics of the liquid distribution obtained by ERT, it was found that particle size has only little impact on induction of pulsating flow and larger particle causes distinct pulses. Smaller particle causes blurred tiny pulses due to higher flow resistance. Larger particle (5 mm) is advantageous for pulsating flow.
Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems
Authors: Zhang Han , Li Fu
Published: February 2014

The traditional solution of the coupled neutronics/ thermal-hydraulics problems has typically been performed by solving the individual field separately and then transferring information between each other. In this paper, full implicit integrate solution to the coupled neutronics/ thermal-hydraulic problem is investigated. There are two advantages compared with the traditional method, which are high temporal accuracy and stability. The five equations of single-phase flow, the solid heat conduction and the neutronics are employed as a simplified model of a nuclear reactor core. All these field equations are solved together in a tightly coupled, nonlinear fashion. Firstly, Newton-based method is employed to solve nonlinear systems due to its local second-order convergence rate. And then the Krylov iterative method is used to solve the linear systems which are from the Newton linearization. The two procedures above are the so-called Newton-Krylov method. Furthermore, in order to improve the performance of the Krylov method, physics-based preconditioner is employed, which is constructed by the physical insight. Finally, several Newton-Krylov solution approaches are carried out to compare the performance of the coupled neutronics / thermal-hydraulic equations.
Photographic Study of Subcooled Pool Boiling Critical Heat Flux on Horizontal Cylinders With Commercial and Rough Surfaces
Authors: Htet Min Han , Fukuda Katsuya , Liu Qiusheng
Published: February 2014

The subcooled pool boiling critical heat flux (CHF) using a 1.2-mm diameter horizontal commercial surface cylinder in water at pressures ranging from atmospheric pressure up to 2063 kPa for a wide ranges of liquid subcoolings from zero to 80 K have been measured to investigate for the mechanisms of subcooled boiling CHF depending on the effects of subcoolings and pressures on the steady and transient CHF. The steady-state CHFs for subcooling were divided into two groups for low and high subcoolings with pressure as a parameter with two different CHF mechanisms resulting from hydrodynamics instability (HI) and heterogeneous spontaneous nucleation (HSN). It was also observed that the measured CHFs resulting from the HI at low pressures in liquid subcoolings as well as the measured CHFs due to HSN at high pressures in high subcoolings respectively on a 1.2-mm diameter of commercial and rough (CS and RS) surface cylinders were approximately same with one another independently of surface conditions. The purpose of the present work is to investigate and compare the heat transfer processes and the vapor behaviors for two mechanisms of subcooled CHFs for the liquid subcooling of 20 K at 101.3 kPa and for the liquid subcooling of 80 K at 690 kPa by mean of photographically. The vapor behavior such as periodic bubble growths and detachments corresponding to HI mechanism at CHF for the liquid subcooling of 20 K at 101.3 kPa and small vapor bubbles existing for some periods on the cylinder corresponding to the HSN mechanism at CHF for the liquid subcooling of 80 K at 690 kPa were investigated.
The Position Design for M310 Reactor Vessel’s Instrumentation Penetration Based on None-Alignment Analysis
Authors: Xia Xin , Du Hua , Li Ning , Xu Bin , Li Yan , Zhao Wei
Published: February 2014

The position of the reactor pressure vessel’s instrumentation penetration is the most important factor that influences the none-alignment between the instrumentation penetration and the instrumentation tube of reactor internals. The more the none-alignment, the more wear the flux thimble will suffer, which will damage the flux detector and cause economic losses. This paper analyzes the none-alignment between the instrumentation penetration and the instrumentation tube, and combines the feedback of the experience in nuclear power plants having been built or being build, puts forward the reasonable design value for the instrumentation penetration’s position.
The Research of Shielding Material for Mixed Field
Authors: Xu Di , Guo Zhen-tao , Li Shu-feng , Zheng Jin-mei , Li Tie-jun , Zhang Long , Hou Xin-peng , Tang Wei , Chen Yan , Zhang Jia
Published: February 2014

This paper developed a kind of highly-filled and ultra-thickness shielding materials; its performance meets shielded neutron-gamma mixed field requirements. Through EDS, theoretical calculation, formulation design, sample preparation and performance test, we carry out the research of base material, neutron absorber, γ-ray absorbent, functional additives and products molding process. The experimental results show that: selection of polyolefin with high hydrogen content as base material, with 15∼25% B4C as a neutron absorber, with 60%∼70% lead powder as a γ-ray absorbent, aiming at the development of protection requirements can make the shielding performance of products reach the ideal level. At the same time, according to the characteristics of system of inorganic filler, the high product thickness, by applying the method of stepped cooling and segmented pressurizing, especially the effective control of crystal formation process to ensure that the product molding and use, overall performance, shielding performance, at the same time improve the yield of the product.
Experimental Study on the Sodium Spray Fire
Authors: Du Haiou , Liu Chen , Wang Rongdong , Hu Wenjun
Published: February 2014

Sodium fire is a design basis accident of sodium-cooled fast reactor. Sodium spray fire experiments were conducted in enclosed cell in this study to understand the sodium spray fire status and how to have an influence on the plant under realistic sodium spray. The process of spray fire experiment was described in the paper. Pressure transducers and thermocouples can be mounted on the room to record pressures and temperatures, respectively. Control and data acquisition system were designed to record most of data for the various instruments. Three experimental results showed realistic sodium spray combustion is mainly in the form of column fire burning. The rise of temperature is drastic on combustion zone. The highest temperature was observed during the sodium spray burning. The experimental results of pressure and temperature were compared to calculation values based on FEUMIX program. It was found the theoretical temperature on 10% sodium atomization of FEUMIX program is coincident with the experimental results of sodium fire burning at over 1500mm combustion zone at time 200 seconds. However, the experimental pressure curve has distinct from theoretical calculation value using FEUMIX program. The results can as a reference to analyze spray sodium fire combustion to help for nuclear safety analysis.
The Use of Risk-Informed Inspection Approaches of Ukraine
Authors: Hryshchenko Bogdan , Polianskyi Mykhailo , Nosovskyy Anatoliy , Sevbo Oleksandr
Published: February 2014

Currently in Ukraine inspection activity is based on deterministic conservative principles, operation experience and expert appraisal of the inspector. Possibility and benefits of PSA are used with low efficiency.

Results of the study conducted by IFC (International Financial Corporation) indicate the absence of a risk-informed approach in the practice of inspections in Ukraine. Also, according to the World Bank appraisal of the investment climate in Ukraine it should be concluded that until regulatory authorities begin implementation of a risk-informed approach in planning inspections in Ukraine random unscheduled events will dominate and won’t provide the goal of State Inspection.

Information which obtained from the PSA helps to direct human and financial resources to the problems research that are the most important for safety, and to eliminate or reduce the requirements, which will reduce expenses of solution for significant issues.

Inspection in the planning, preparation, implementation and evaluation of the results of which, in addition to deterministic estimates, operating experience and expertise evaluating risk are used is called the risk-informed inspection. The use of risk-informed approaches allows to: focus inspections on design and operational aspects, which have dominant influence on the safety of nuclear power plants (NPP); improve the schedule of inspections (recording of risk assessments in determining the scope, frequency, and type of inspection); and use an additional source of information on the systems and the components of power units, personnel availability.

Applying a risk-informed approach to inspection, inspectors can focus primarily on systems that make the largest contribution to core damage frequency, failure of which leads to significant increase of CDF. Risk-informed approach allows to select the most important elements to test systems that will increase the efficiency and quality of inspections.

Based on the above it can be concluded that the experience of inspector, his knowledge of the power unit design, the process, the mechanisms of failure of equipment and of accident running, the use of information on the importance of components and systems for the safety of nuclear power plants, obtained from PSA — an effective way to achieve the best results in improving safety.
Experimental Modeling of Light Phase Effect on Heat Transfer in Rod Bundle
Authors: Kashinsky O. N. , Lobanov P. D. , Kurdyumov A. S. , Pribaturin N. A. , Volkov S. E.
Published: February 2014

Experiments in two-phase gas-liquid flow in a vertical bundle of rods were performed. The void fraction distribution and heat transfer from heated central rod were considered. Comparisons for cases of single phase and two-phase flows are presented. Gas addition to the flow results in heat transfer enhancement. The position of gas injection point plays a significant role on heat transfer characteristics. A non uniform gas phase distribution around the central rod of the bundle was obtained.
Experimental Investigation of Non Stationary Heat Transfer Between Fluid and Solid Body
Authors: Marcinkiewicz Jerzy , Taler Jan , Cebula Artur
Published: February 2014

The presented work is a result of actions taken in connection to analyses of root cause of damages (cracks and one brake) in the control rod shafts in Swedish BWR Forsmark 3. The damages were detected during the refueling outage 2008. It has been found that damages were caused by thermal fatigue. Extensive analyses of flow and temperature fields around the shaft were performed using transient CFD calculations [1, 2]. The character of the fluctuating thermal loadings on the shaft was confirmed by a limited experiment [3]. However the CFD-calculations of heat transfer between the water and the shaft have not been validated experimentally. In order to validate CFD-calculations of heat transfer between the water and the solid body the measurements of the non-stationary heat transfer are planned. The paper presents the method of determination of heat flux and temperature on the surface of the body based on temperature measurements at some discrete points beneath the surface and solving the inverse heat conduction problem (IHCP). Software was developed for performing measurements and calculations. Main parts of measurement system particularly design and manufacturing of measuring items, thermocouple installation, construction of test stand for initial testing and calibration are described. Verification of thermocouple locations was performed using computer tomography and the actual locations were introduced in to the calculation algorithm in order to improve accuracy of heat flux determination. Heat flux measuring error has been determined based on assumed random error in temperature measurement and accuracy of location verification (computer tomography). Results of initial verifying tests are presented and discussed. The measuring system is now ready for performing measurements of transient heat transfer in configurations that can occur in a reactor environment.
Analysis and Evaluation of the Setting Value Drift of SEBIM Relief Valve Used in Nuclear Power Plant
Authors: Yang Xiaochuan , Luo Chunlei , Qu Meng , Shen Rongfa , Che Yinhui
Published: February 2014

SEBIM relief valve is mainly used in safety important system of the nuclear power plant to provide the overpressure protection for the pressure vessel and pressure pipeline which related to the nuclear safety. It is probable that the system will lose the overpressure protection function and lead to a major security hidden danger in the nuclear power plant if the relief valve setting value happens drift. So the problem of the SEBIM relief valve setting value drift has becoming the focus of attention of the whole nuclear industry.

It is of great significance to study the setting value drift condition of SEBIM relief valve in the practical application. This paper analyzes the structure and operating principle of the SEBIM valve, and studies the main failure mode of the setting value drift of SEBIM relief valve combining with the operating environment of the equipment. And this paper collects the setting value check data of the SEBIM valve used in two domestic nuclear power plants (These two power plants are named NPP A and NPP B in this paper) during its operation period, and carries on the analysis and evaluation.
Research and Development of Self-Priming Venturi Scrubber for Filter Venting: Preliminary Analysis and Observation of Hydraulic Behavior in Venturi Scrubber
Authors: Horiguchi Naoki , Yoshida Hiroyuki , Uesawa Shin-ichiro , Kaneko Akiko , Abe Yutaka
Published: February 2014

As revealed by Fukushima Daiichi nuclear disaster, countermeasures against severe accident in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing the diffusion of radioactive materials, it is important to develop the device which allows filtered venting of contaminated high pressure gas. In the filtered venting system that used in European reactors, so called Venturi scrubbers are used to realize filtered venting without any power supply. The Venturi scrubber operates without a power supply of high pressure gas filled into the containment. In this apparatus, scrubbing of contaminated gas is promoted by both gas releases through a submerged Venturi tube which is one of the major components of Venturi scrubber and liquid splay flow formed by liquid suction through a hole for suction provided at the throat part of the Venturi tube. This type of Venturi scrubbers is called self-priming ones. However, the mechanism of a self-priming Venturi scrubber including effects of gas flow rate and shape of Venturi tube is understood insufficiently in the previous studies.

In this study, to provide knowledge about the detailed mechanism of filtering and evaluation method for operating condition of the filtered venting system with Venturi scrubber, both experimental and numerical research works have been performed. In this paper, as a previous step of designing and making a self-priming Venturi tube, hydraulic behavior in a rectangular Venturi tube is taken by high speed camera, by the air-water experiment under atmospheric condition to check the operation in a self-priming mode and the validity of the experimental method and the visualization technique. And numerical analysis of a circular Venturi tube is conducted as a preliminary analysis, by TPFIT (Two-Phase Flow simulation code with Interface Tracking) to predict its operation. As results, the Venturi scrubbers used in experiment and simulated in numerical analysis were operated in a self-priming mode. By observed results of the hydraulic behaviors in the Venturi tube, the validity of the experimental method and the visualization technique were confirmed. And it is expected that the simulated circular Venturi scrubber in this numerical simulation was operated in a self-priming mode in a real situation.
CFD Simulation of Flashing Boiling Flow in the Containment Cooling Condensers (CCC) System of KERENA™ Reactor
Authors: Liao Yixiang , Lucas Dirk , Krepper Eckhard , Rzehak Roland
Published: February 2014

The flashing boiling phenomenon occurring inside the Containment Cooling Condensers (CCC) passive system of KERENA™ reactor was simulated by using the CFD (Computational Fluid Dynamics) code ANSYS CFX. The steam generation was assumed to be caused by thermal phase change, which is controlled by interphase heat transfer. In general, the model setup is able to capture the generation and disappearance of steam and the fluctuation of pressure and temperature observed during the flashing. However, since the nucleation and bubble growth process was neglected in the simulation, the initialization of evaporation was found to be a key issue, which depends on the prescribed bubble size sensitively. In order to get a satisfying agreement with the experimental data, a relatively large size has to be assumed for the selected test case (db>10 mm). The CFD results are qualitatively comparable to those obtained by using the system code ATHLET, except that the mass transfer rate is higher and the fluctuation frequency and altitude is lower. In addition, the assumption of constant bubble size might introduce significant error. The change of bubble size as well as the effect of nucleation and bubble growth should be investigated in the future work. For this purpose, however, further experimental data such as evaporation rate, phase and bubble size distribution, phase velocity are required.
Calculation of External Radiation in Dry Spent Nuclear Fuel Storage Container
Authors: Zaiets V.
Published: February 2014

Zaporozhye Nuclear Power Plant (NPP) commissioned a Spent-Fuel Dry-Storage Facility (SFDSF) with a 50-year service life. The SFDSF design is based on the proven technology of the US Duke Engineering & Services Company and meets international standards, rules and regulations on nuclear and radiation safety [1].

The objectives of this paper are calculation of the external neutron and γ-radiation of a single container, and prediction of radiation characteristics during a storage period up to 50 years.

The calculations were made for VSC-24 special ventilated containers manufactured by Sierra Nuclear Corporation (USA) [2].

The analysis shows that the change in gamma energy spectrum results in a faster decrease in dose with time which is in agreement with observations in the SFDSF. The γ-radiation dose rate is mainly composed of 0.5 to 2.5 MeV quanta.
Study on Condensation Heat Transfer on Micro Structured Surfaces (Effect on Condensation Heat Transfer of Metal Sputtering Surfaces)
Authors: Yamazaki Kohei , Ohtake Hiroyasu , Hasegawa Koji
Published: February 2014

The present study was intended to examine how the condensation heat transfer, especially the dropwise condensation, was affected by modifying the surface nature. In the present study, condensation heat transfer experiments for steam were performed by using mirror-finished copper surface and some very thin metal-film surfaces by using sputtering on mirror-finished copper block. That is, the effects on pattern of condensation heat transfer, i.e., dropwise or film-wise condensation, of metal-sputtered surfaces were examined experimentally and qualitatively. The present experimental results showed that the condensation on sputtered metal surfaces of Copper (Cu), Chromium (Cr) and Lead (Pb), became dropwise condensation. The heat transfer coefficients were ten times higher than the Nusselt equation. The condensation on sputtered metal surface of Titanium (Ti) became filmwise condensation. High contact angle was trended to be dropwise condensation on very thin metal-film surfaces by using sputtering.