bluuebunny/crossref_metadata_2025 huggingface dataset:
Creep Stress Analyses Affected by Defect Geometries on P91 Pipe With Local Wall Thinning Under High Temperature
Authors: Xue Jilin , Zhou Changyu , Peng Jian
Published: April 2011

P91 heat-resistant steel pipes are widely used at high temperature in power plants and nuclear power plants. The service condition and manufacturing process may produce defects of local wall thinning, which may result in stress redistribution of the pipes during the service process at elevated temperature. For the purpose of understanding the creep stress and strain accumulation affected by local wall thinning geometries under creep condition, three groups of models were calculated, using three-dimensional models based on finite element analyses (FEA) codes ABAQUS. In this study, monotonic internal pressure was conducted on P91 full-scale steel pipes at 625°C, with local wall thinning located at the inner surface. Then, the creep strain and stress of pipes after 100,000h could be obtained corresponding to different models. Based on the analysis, the figures of creep stress and strain varying with defect geometries were plotted. Then, the stress and strain of pipes with local wall thinning affected by defect geometries, including different defect depths, different defect axial lengths and different defect hoop angles, were discussed. The results indicate that creep stress and creep strain increase with defect geometries. The variation laws have been summarized. The research results can provide the possibility on safety assessment and structure integrity analysis of the pipe with local wall thinning at high temperature effectively.
Study on a Methodology of Human Factor Engineering Operating Experience Review for Nuclear Power Plant
Authors: Gu Danying , Zhang Shuhui , Ning Zhonghe
Published: April 2011

The reviewing of operating experience at nuclear power plants (NPP) is not only critically important to safe and reliable operations, but also useful to guide the design of new plants which are similar to the current one under review. How to identify and analyze the safety-related operating experience and then implement a more extensive review is a vital and challengeable issue. In this paper, a methodology of human factor engineering (HFE) operating experience review (OER) is proposed for NPP. The need for the application of HFE in the life cycle activities of NPP and other nuclear facilities has been demonstrated by plant operating histories and regulatory and industry reviews. As a very important element of HFE, the OER is performed from the beginning of the design process. The main purpose of performing an OER is to verify that the applicant has identified and analyzed HFE-related safety problems and issues in previous designs that are similar to the current one. In this way, negative features associated with predecessor designs may be avoided in the current NPP design while retaining positive features. The research of OER concentrates on the aspect of review criterion, scope and implementation procedure of the HFE-related operating experience. As the NRC requirement, the scope of operating experience can be divided into six types in accordance with sources of information. The implementation procedures of USA and China are introduced, respectively. The resolution of HFE OER issues involve function allocation, changes in automation, HSI equipment design, procedures, training, and so forth. The OER conclusions can contribute to other HFE activities and improve the safety, reliability and usability of the HSI design in NPP.
Advances and Research Needs in Risk-Informed In-Service Inspection Methodology
Authors: Yuan X.-X.
Published: April 2011

While risk-informed in-service inspection (RI-ISI) program has been applied in several countries to enhance the traditional periodic inspection program (PIP), many other countries are waiting for more successful implementation experiences to be accumulated. Canadian Nuclear Safety Commission (CNSC), the regulatory body of nuclear industry in Canada, became increasingly interested in the risk-informed decision making methodology. Several small-scale pilot studies on RI-ISI have been initiated by Canadian utilities during the past few years. Nevertheless, a RI-ISI methodology appropriate for the CANDU technology that can be accepted by the stakeholders has yet to be developed. The development of the RI-ISI methodologies derived from the PWR/BWR operating experiences is first reviewed, followed by an examination of Canadian periodic inspection standard CSA N285.4 and its evolution from a RI-ISI perspective. Finally several key technical issues and research needs in developing an advanced RI-ISI methodology for nuclear power plants are identified.
Image Analysis of Carry-Under by Falling Liquid Flow
Authors: Ohtsuka Tomohiko , Haraguchi Naoki , Ohtake Hiroyasu , Koizumi Yasuo
Published: April 2011

Bubble carry-under into the water pool was examined. In experiments, a water jet from a nozzle of 5 mm in diameter plunged into the water pool. The distance between the nozzle outlet and the pool surface was 246 mm. Flow behavior in the water pool and also the state of the water jet surface were recorded with a high speed video camera. Following conclusions were obtained. When the flow rate of the water jet was small, the water jet disintegrated into small drops on the way from the nozzle outlet to the pool surface. The wave appearing position moved downward as the flow rate was increased. When the wave length reached the Kelvin-Helmholtz critical wave length, the water jet disintegrated into drops. When flow rate of the water jet was increased, the surface of the water jet became smooth and no perturbation was observed. The carry-under was not observed in this situation. When the flow rate of the water jet was further increased, large waves came to appear on the water jet surface. The wave appearing position moved upward as the flow rate was increased. Even if the wave length on the water jet reached the Kelvin-Helmholtz critical wave length, the water jet did not disintegrate into drops and the water jet plunges into the pool with large waves on the water jet. The penetration depth in this case was deep and the volume of the bubble carry-under was large compared with the case that the water jet disintegrated into drops.
Study on Corrosion Control in Reactors Using Radiation Induced Surface Activation (RISA): Mechanism Behind Stainless Steel Durability Due to RISA Against Crevice Corrosion
Authors: Mabuchi Shogo , Hazuku Tatsuya , Motoda Shin-ichi , Takamasa Tomoji , Uematsu Susumu , Furuya Mashiro
Published: April 2011

This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel in piping for nuclear reactors. This employs an effect of Radiation Induced Surface Activation (RISA). The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures.
Computational Analysis of Pressure Loading and Impulse Distribution in the Reactor Cavity During the Process of Ex-Vessel Steam Explosion
Authors: Huang Xi , Wang Xi , Yang Yan-hua
Published: April 2011

In order to investigate the magnitude and distribution of pressure and impulse in the reactor cavity during the process of fuel-coolant interaction includes both the premix and steam explosion, a three dimensional multiphase fluid simulation code MC3D has been used based on the structural and environmental characteristic of the sever accident state of 1000 MW level PWR Linao Nuclear Power Station Phase II with the consideration of both pressure variation and impulse variation. The results indicate that the process of premixing has a low level pressure variation with a long duration and the affect of this process is negligible. On the contrary, explosion stage creates a instant and sharp pressure pulse which threatens the integrity of the reactor cavity structure. In addition, the mass of the melting drop contacts the coolant all over the calculating grids has a significant effect regarding the magnitude of the impulse created by steam explosion.
Photographic Study on Two-Phase Flow Patterns of Water in a Single-Side Heated Narrow Rectangular Channel
Authors: Wang Jun-Feng , Huang Yan-Ping , Wang Yan-Lin
Published: April 2011

Visualized experimental observation on flow patterns during flow boiling of water under single-side heated and fluid-inlet subcooled conditions in a vertical narrow rectangular channel with the cross-section of 40×3mm2 have been carried out. Four discernible flow patterns which names dispersed bubbly, coalesced bubbly, churn flow and annular flow are obtained. Flow visualization in two dimensions of two-phase flow patterns for narrow rectangular channel, which provided clearer evidence to distinguish flow patterns, have been performed. Based on the experimental results, a flow pattern map for single-side heated narrow rectangular channel has been developed and then compared with the exiting maps and flow transition criteria.
Nonlinear Analysis for a Two-Phase Natural Circulation Loop Under Low-Pressure
Authors: Kazem Arrdaneh , Salman Zaferanlouei , Farahi Mohsen , Ahmadi Asad Allah
Published: April 2011

The purpose of this paper is to develop a nonlinear model to investigate the instabilities of a two-phase natural circulation loop under low-pressure condition. Inlet velocity oscillations and the corresponding trajectories are respectively presented in the time evolution planes and phase planes. We obtain a stability map to explore the instability regions of this natural circulation loop. The results show that the considered loop has two unstable regions, instability type-I in the low power region and instability type-II in the high power region. Then the parametric study is carried out to understand the relation between the parameters of system and two types of instability. The parametric study reveals that lengthening the riser has an unstable effect on system stability. Thus, lengthening the riser causes a reduction in the stability region in the both low power and high power levels. Also it can be observed that by increasing the form loss coefficient at the inlet of heated section or in the downcomer section, the stability region expands, however by increasing the form loss coefficient at the outlet of heated section or in the upper horizontal section, the stability region decreases consequently.
Experimental Study of Bubble Behavior in a Two-Dimensional Particle Bed With High Solid Holdup
Authors: Cheng Songbai , Hirahara Daisuke , Tanaka Youhei , Gondai Yoji , Matsumoto Tatsuya , Morita Koji , Fukuda Kenji , Yamanao Hidemasa , Suzuki Tohru , Tobita Yoshiharu
Published: April 2011

In a core disruptive accident of a fast breeder reactor, the post accident heat removal is crucial to achieve in-vessel core retention. Therefore, a series of experiments on bubble behavior in a particle bed was performed to clarify three-phase flow dynamics in debris bed, which is essential in heat-removal capability, under coolant boiling conditions. Although in the past several experiments have been carried out in the gas-liquid-solid system to investigate the bubble dynamics, most of them were under lower solid holdup (≤ 0.5), where the solid-phase influence may be not so important as much as the liquid phase. While for this study, the solid holdup is much higher (> 0.55) where the particle-bubble interaction may be dominated. The current experiment was conducted in a 2D tank with the dimensions of 300 mm height, 200 mm width and 10 mm gap thickness. Water was used as liquid phase, while bubbles were generated by injecting nitrogen gas from the bottom of the tank. Various experimental parameters were taken, including different particle bed height (from 30 mm to 200 mm), various particle diameter (from 0.4 mm to 6 mm), different particle type (acrylic, glass, alumina and zirconia beads), and different nitrogen gas flow rate (around 1.75 ml/min and 2.7 ml/min). By using digital image analysis method, three kinds of bubble rise behavior were observed under current experimental conditions and confirmed by the quantitative detailed analysis of bubble rise properties including bubble departure frequency and bubble departure size. This experiment is expected in the future to provide appropriate quantitative data for the analysis and verification of SIMMER-III, an advanced fast reactor safety analysis code.
Gas Flow Simulations in Randomly Distributed Pebbles
Authors: Zhao Xiang , Montgomery Trent , Zhang Sijun
Published: April 2011

In this paper, computational fluid dynamics (CFD) gas flow simulations are carried out. In CFD calculations, geometry modeling and physical modeling are crucial to CFD results. The effects of the treatments of the inter-pebble contacts on gas flow fields and heat transfer are examined; a sensitivity analysis for the gap size is conducted with two spherical pebbles, in which the inter-pebble region is modeled by means of two types of inter-pebble gaps and two kinds of direct contacts. On the other hand, both of large eddy simulation (LES) and Reynolds-averaged Navier-Stokes (RANS) models are employed to investigate the turbulent effects. It is found that the flow fields and relevant heat transfer are significantly dependent on the modeling of the inter-pebble regions. The calculations indicate the complex flow structures present within the voids between the fuel pebbles.
Three Types of a Passive Safety Containment for a Near Future BWR With Active and Passive Safety Systems
Authors: Sato Takashi , Akinaga Makoto , Kojima Yoshihiro , Murakami Tsunekazu , Hosomi Kenji
Published: April 2011

The paper presents three types of a passive safety containment for a near future BWR. They are tentatively named Mark S+, Mark D and Mark X containments in the paper. They all have a leak tight secondary containment vessel (SCV) in order to meet the reactor site criteria without relying on an active standby gas treatment system at a DBA LOCA. One of their common features is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The containment pressure can be limited within the design pressure. Even if a large amount of hydrogen is generated at a severe accident, it can be released into the SCV. Hydrogen detonation or deflagration is completely prevented without using igniters. Another feature is the capability to submerge the PCV and the RPV above the core level without relying on accident management. The core debris is completely submerged not only ex-vessel but also in-vessel. The third feature is robustness against external events such as a large commercial airplane crash. All the containments have built-in passive safety systems (BIPSS) including a passive containment cooling system (PCCS) and a passive cooling core catcher that has radial cooling channels. The Mark S+ and Mark D containments are applicable to a large power BWR up to 1830 MWe. The SCV is made of steel-concrete composite. The PCV can be vented into the inerted part of the SCV at a severe accident. The Mark X containment has the steel secondary containment vessel (SSCV) and can be cooled by natural convection of outside air. It can accommodate a medium power BWR up to about 1000 MWe and has a permanent grace period without replenishing the PCCS pool. In all cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides in-depth protection against severe accidents and also enables N+2 design. All the three containments coupled with the IDHS can potentially provide an evacuation free plant at a severe accident caused by severe natural disasters such as a giant earthquake, a tsunami, a mega hurricane, and so on.
Visualization of Bubble Growth and Departure in a Vertical Narrow Rectangular Channel
Authors: Xu Jian-Jun , Chen Bing-De , Wang Xiao-Jun
Published: April 2011

It is very important to study bubble growth and departure from the nucleation site for better understanding of boiling heat transfer in a narrow channel. Bubble growth and departure in a narrow rectangular under atmosphere pressure is visually observed by the wide and narrow side of the narrow rectangular channel using high speed digital camera. There is a small bubble contact diameter between the bubble base and heating surface when the bubble is growing at the nucleation site, and the growing bubble shape is almost spherical. The bubble growth law at the different nucleation sites is almost uniform under the condition of the same thermal parameters, but bubble departure diameters are obvious distinct because of different sizes of nucleation sites. In the current study, the bubble growth rate in a narrow rectangular channel is small, and the bubble departure time is long, the bubble growth diameter can be predicted by using the amendatory Zuber expression. The effect of thermal parameters on the mean bubble departure diameters is statistical analysed in the view window, the mean bubble departure diameters decrease with increasing heat flux, the mean bubble departure diameters decrease with increasing inlet subcooling, the mean bubble departure diameters decrease with increasing bulk flow velocity.
Study on HSE Management System of AP1000 Nuclear Plant Project
Authors: He Xiaogang , Fang Xiaopeng , Wei Zhong , Yang Yanli , Sun Zhiyou
Published: April 2011

Being a third generation advanced nuclear plant type, AP1000 has the characteristics of modular construction, “Open-Top” method lifting, etc, which simultaneously present a new challenge for site HSE management. Through studying Health, Safety & Environment (HSE) site management of the first AP1000 nuclear plant (ANP) under construction, this research analyzes the difference of HSE management aspects, such as management commitment and responsibility, HSE awareness and concept, training and education, HSE reward & punishment system and other aspects between China and the west. It also puts forward the specific idea for developing site HSE management system suitable for ANP technology characteristic and China’s actual conditions, which will provide a guarantee for the safe and smooth construction of ANP project.
New Human-Machine Interface at VR-1 Training Reactor
Authors: Kropi´k Martin , Rataj Jan , Jurˇicˇkova´ Monika
Published: April 2011

The paper describes a new human-machine (HMI) interface of the VR-1 nuclear training reactor at the Czech Technical University in Prague. The VR-1 reactor is primarily used for training of university students and future nuclear power plant staff. The new HMI was designed to meet functional, ergonomic and aesthetic requirements. It contains a PC with two monitors. The first alphanumerical monitor presents text messages about the reactor operation and status; next, the operator can enter commands to control the reactor operation. The second graphical monitor provides parameters of reactor operation and shows the course of the reactor power and other parameters. Furthermore, it is able to display the core configuration, perform reactivity calculations, etc. The HMI is also equipped with an alarm annunciator. Due to a high number of foreign students and visitors at the reactor, the Czech and English language versions of the user interface are available. The HMI contains also a History server which provides a very detailed storage and future presentation of the reactor operation. The new HMI improves safety and comfort of the reactor utilization, facilitates experiments and training, and provides better support for foreign visitors.
Thermal Loading Failure Analysis of Molten Pool in the Reactor Vessel
Authors: Yang Xiao , Yang Yanhua , Kuang Bo , Fu Xiaoliang
Published: April 2011

Uncertainties are addressed in the special context of assessing and managing risks from rare, severe-consequence hazards. Risk Oriented Accident Analysis Methodology (ROAAM) is used to analyze uncertainties during severe accidents analysis in nuclear power plants. In-vessel Retention (IVR) is one of the mitigations for severe accidents which will cause core damage. By external reactor vessel cooling (ERVC), the integrity of the reactor vessel is preserved. The success criterion for IVR is the local heat flux on the wall of lower head is less than the critical heat flux (CHF). This paper analyzes the uncertain parameters which decide the mitigation to be successful or fail. Two bounding structures and 4 molten pool steady states are defined. And the success probability of IVR is evaluated with a molten pool heat transfer model. Then the effectiveness of IVR-ERVC under the two bounding structures is evaluated.
Adaptive Unbalance Vibration Control of Active Magnetic Bearing Systems for the HTR-10GT
Authors: Hui Chao , Shi Lei , Wang Jie , Yu Suyuan
Published: April 2011

An Adaptive Unbalance Vibration Control (AUVC) method is studied based on the flexible active magnetic bearing-rotor experimental system AMB-PII, which is built up to simulate the helium turbo compressor rotor of the 10MW high temperature gas-cooled reactor with direct cycle (HTR-10GT) in compliance with the dynamic similarity principle. By constructing the required frequency compensation signal automatically through different feedforward approaches, two types of control methods are realized: the “displacement nulling” control which can effectively cancel the rotor vibration amplitude within the system bandwidth, and the “current nulling” method which can let the rotor rotate around its inertia axis and eliminate the disturbances of the currents in the electromagnet windings to significantly attenuate the mechanical vibration. The simulation and experimental results have proved that this kind of AUVC integrating feedforward algorithm can achieve good performance improvements, which will provide good experience and design references for the future application of magnetic bearings in the HTR-10GT project.
Signal Processing Schemes for Defect Identification and Sizing From Signal of Eddy Current Testing in Nuclear Power Plant
Authors: Tian Shuxia , Chen Zhenmao , Zhang Dongli , Geng Qiang
Published: April 2011

In this paper, signal processing schemes developed for upgrading the Eddy Current Testing (ECT) system utilized for In Service Inspection (ISI) of Steam Generator (SG) tubes of a nuclear power plant are presented. Schemes for the wobbling noise recognition, signal indication extraction, signal classification, and the separation of mixed signals are proposed. The schemes are applied to both signals measured in a SG mock-up and signals from short tube test-pieces measured in a laboratory environment. It is demonstrated that the methods work well even for mixed signals in the support plate region of SG tubes. As it is possible to adjust the position of a signal point based on the signature of the support plates, noise signals from a welding line or a bending zone can be recognized based on the information of design and manufactory. The mixed signals of defects and support structures are processed by using a similarity analysis strategy and the identification of defect signals is performed by means of both a neural network approach and a statistic method. Satisfactory results are obtained for the processing of measured signals.
The Outline of the Five-Percent Power Uprate Project in Tokai-2
Authors: Ohata Hitoshi , Nishibata Toshikazu , Onose Tetsuya
Published: April 2011

Reactor thermal power uprate (Power uprate) of operating light water reactors has long successful experiences in many nuclear power plants in the United States of America and European countries since late 1970’s. And it will be also introduced in Japan soon. This paper mainly describes the outline of the attempt of five-percent reactor thermal power uprate of Tokai No.2 Nuclear Power Station (Tokai-2) operated by the Japan Atomic Power Company (JAPC). It will be the leading case in Japan. Tokai-2 is GE type Boiling Water Reactor (BWR) of 1100 MW licensed electric power output and it commenced commercial operation in November 28, 1978. Power uprate is an effective approach for increasing electric power output. And it is recognized as one of the measures for effective and efficient use of existing Japanese operating nuclear power plants. It can contribute to inexpensive and stable electric power supply increase. Especially “Stretch Power Uprate (SPU)” requires only minor equipment modification or component replacement. It is also a countermeasure against global warming. Therefore it is a common theme to be accomplished in the near future for both Japanese electric power companies and government. JAPC started feasibility studies on power uprate in 2003. And in 2007, JAPC established a plan to achieve five-percent power uprate in Tokai-2 and announced this project to the public. This is a leading attempt in the Japanese electric power companies and it is the first case under the current Japanese regulatory requirements. In this plan, JAPC reflected lessons learned from preceding nuclear power plants in the United States and European countries, and tried to make most use of the performance of existing systems and components in Tokai-2 which have been periodically or timely renewed by utilizing more reliable and efficient design. JAPC plans to submit application documents to amend current License for Reactor Establishment Permit shortly. It will contain a complete set of revised safety analysis results based on the uprated reactor thermal power condition. Successful introduction of Tokai-2 power uprate will contribute to the establishment of regulatory process for power uprate in Japan and following attempts by other Japanese electric power companies.
Experimental and Computational Studies of LNPP-2 Passive System for Severe Accident Management
Authors: Sidorov Valery G. , Bezlepkin Vladimir , Alekseev Sergej , Semashko Sergey , Ivkov Igor , Kukhtevich Vladimir
Published: April 2011

The project of nuclear station LNPP-2 with a reactor power plant VVER type by electrical power 1200 MVt involves a number of new design solutions to increase of parameters of safety. The passive containment heat removal system and heat removal system via steam generators is including of number of such solutions. Passive heat removal system via steam generators (PHRS/SG) is assigned for remove of residual heat of reactors core to final heat absorber (atmosphere) through a secondary circuit at DEC accident. The system PHRS/SG duplicates cooling-down system via SG to final heat absorber in case of impossibility of realization of its design functions. Containment heat removal system (PHRS/C) is assigned for remove of residual heat from containment in accidents with heat-transfer emissions from primary circuit. PHRS/C duplicates functions of a spray system to reduce of pressure under containment in case of spray system failure. In the substantiation of passive security systems the complex in SPbAEP of computational and experimental analysis was executed, the main results of which are shown in the present report.
Analysis of Orifice Plate Differential Pressure Noise in Horizontal and Vertical Rising Gas-Water Two-Phase Flow
Authors: Li Xianfa , Zhong Shuoping , Sun Yanfei
Published: April 2011

It is an important achievement of modern techniques to determine the mass flow rate and the phase fraction of wet steam by measuring the orifice plate differential pressure noise. The orifice plate differential pressure noise of air-water two-phase flow in horizontal and vertical rising pipelines were analyzed. Kinds of calculation methods were tried to get the differential pressure noise. From the difference waveform of the differential pressure square root that the acquisition card got and the mean square root of the sample that got before, the first in first out (FIFO) principle was used to get the differential pressure noise. Result shows that the differential pressure noise has different level at different vapor flow rate with the same water flow rate, conclusions show that the two-parameter measurement by using orifice plate differential pressure noise may be possibly used in vertical rising gas-water two phase flow.
Extraction of Actinides and Lanthanides by Supercritical Fluid
Authors: Zhu Liyang , Duan Wuhua , Xu Jingming , Zhu Yongjun
Published: April 2011

Reprocessing of used nuclear fuel and treatment of nuclear waste are important issues for the sustainable development of nuclear energy. It is necessary to develop novel nuclear waste treatment technologies to meet the goal of minimizing the secondary liquid waste. Supercritical fluids are considered green solvents in chemical engineering process. It gains growing interest to treat nuclear waste using supercritical fluid extraction recently, because it can greatly decrease the secondary liquid waste with high radioactivity. During the past two decades, extraction of actinides and lanthanides by supercritical fluid has been intensively studied in some countries, and many important progresses have been made. However, the prospect of industrial application of supercritical fluid extraction technology in reprocessing of used nuclear fuel and treatment of nuclear waste is still unclear. In this paper, extraction of actinides and lanthanides from various matrixes or from their oxides by supercritical fluid including the experimental results, extraction mechanism and kinetic process was reviewed. The engineering demonstration projects were introduced. The trend of industrial application of supercritical fluid extraction technology in nuclear waste management was also discussed.
The Treatment Room and Maze Design for Medical Accelerators
Authors: Jing Shang , Li Juexin , Li Yuxiong
Published: April 2011

This paper analyses the shielding design of medical accelerators. To find the best plan among them, which need the lowest cost to obtain the same effect, some theoretical calculations according to the recommended methods of NCRP 151 report have been done. The paper also compares the effects on dose equivalent at the maze door when beam direction is parallel or perpendicular to the maze wall separately. Then it concludes that the photon dose at the maze door will be lower by a factor of one magnitude if the door locates at the side of the maze but not at the terminal of it. In addition, the neutron capture γ-ray and the photo neutron dose will also be greatly reduced in this situation. It should be pointed out that the beam directly hitting the maze wall is not a recommended design, although it can meet the final dose standard. Since more and more medical accelerators are used in radiotherapy, this work may be helpful to the medical accelerator shielding design both for maze and treatment room.
Corrosion Properties of Welded Ferritic-Martensitic Steels in Liquid Lead-Bismuth at 600C
Authors: Mustari Asril Pramutadi Andi , Takahashi Minoru
Published: April 2011

A static corrosion test for ferritic-martensitic steel HCM12A with three types of welding in lead-bismuth eutectic (LBE) was conducted at 600°C for 500 h. The specifications of TIG welding, YAG laser welding and electron beam welding were 40A/12V of electrical conductance, 278W/14.3J of heat input and 5.0 mA of beam electric current, respectively. Oxygen concentration was controlled to be about 4.7×10−6wt% by injecting Ar-H2-H2O gas mixture into LBE. After corrosion test, cross sections of the specimens were analyzed by using optical microscope and SEM/EDX. Toughness profiles of fusion zone and base metal were determined by using Vickers hardness tool. The results show coarse grain structure in fusion zone (FZ) and typical ferritic-martensitic grain structure in base metal (BM). It was found that those Cr-rich spinel oxide layer and diffusion zone were formed on the surface. The total thickness of oxide layer and diffusion zone on and in the fusion zone was about 18–30μm. It was much thicker than 10–15μm on BM. The hardness was higher in the fusion zone than in the BM.
Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4
Authors: Ye Xuhua , Peng Minjun , Liu Jiange
Published: April 2011

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.
Comparison of Three-Rod Bundle Data With Existing Heat-Transfer Correlations for Bare Vertical Tubes
Authors: King Krysten , Farah Amjad , Gupta Sahil , Mokry Sarah , Pioro Igor
Published: April 2011

Many heat-transfer correlations exist for bare tubes cooled with SuperCritical Water (SCW). However, there is very few correlations that describe SCW heat transfer in bundles. Due to the lack of extensive data on bundles, a limited dataset on heat transfer in a SCW-cooled bundle was studied and analyzed using existing bare-tube correlations to find the best-fit correlation. This dataset was obtained by Razumovskiy et al. (National Technical University of Ukraine “KPI”) in SCW flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within a range from 800 to 3000 kg/m2s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2. The objective of this study is to compare bare-tube SCW heat-transfer correlations with the data on 1- and 3-rod bundles. This work is in support of SuperCritical Water-cooled Reactors (SCWRs) as one of the six concepts of Generation-IV nuclear systems. SCWRs will operate at pressures of ∼25MPa and inlet temperatures of 350°C.
A Core Catcher Concept and First Experimental Results
Authors: Kim Hwan Yeol , Ha Kwang Soon , Kim Jong Hwan , Hong Seong Wan , Song Jin Ho
Published: April 2011

In a postulated core melt accident, if a molten core is released outside a reactor vessel despite taking mitigation actions, the core debris would relocate in the reactor cavity region and attack the concrete wall and basemat of the reactor cavity. This will potentially result in inevitable concrete decompositions and possible radiological releases. To prevent direct contact of the melt and basemat concrete of the cavity, a core catcher concept is suggested, which can passively arrest and stabilize the molten core material inside the reactor cavity. The core catcher system includes a retention device for the molten core material, a cooling water storage tank, and a compressed gas tank. Upon ablation of the sacrificial layer on top of the retention device while molten core material is discharged, a mixture of water and gas is injected from below. It is expected that a simultaneous injection of water and gas could prevent a possible steam explosion/spike. It could also suppress the rapid release of steam which might result in fast over-pressurization of the containment. A test facility for the core catcher using a thermite reaction technique for the generation of the melt was designed and constructed at KAERI. The first series of tests were performed by using a mixture of Al, Fe2O3, and CaO as a stimulant. As a first try, only water was injected from the bottom of the melt through five water injection nozzles when the melt front reached the water injection nozzles. In this paper, the core catcher concept and the related provisions are suggested. A description of the test facility for the core catcher, the thermite composition, and the methods of experiment is included. The first experimental results with only water injected from the bottom of the melt are discussed.
Development of Nuclear-Specific Coriolis Flowmeters Based on the Straight-Tube Technology
Authors: Wang Tao , Hussain Yousif
Published: April 2011

The application of Coriolis flowmeters to various industries has been expanding since they were successfully introduced in the market some 30 years ago. This paper specifically reports the development of Coriolis flowmeters for the nuclear industry based on the latest straight-tube technology. An advanced numerical method combining the strength of commercially available simulation packages and in-house theoretical work was used in the development. This includes the initial prediction of flow sensitivity using the fluid-structure interaction theory for straight flow tubes to ensure sufficient measurement accuracy under various process conditions. It also includes further detailed analysis of the entire flowmeter to make sure its natural frequencies away from the seismic frequency range due to nuclear application safety reasons. Furthermore, a new calibration and testing procedure is reported in this paper, which includes the normal calibration condition and also simulates possible process conditions during the nuclear applications (e.g. varying fluid temperature). Additionally, calculations based on the Design by Formula approach according to the governing codes, particularly ASME Boiler and Pressure Vessel Code, Section III, are also reported. These simplified calculations are of significant importance for Coriolis flowmeters to be applied to nuclear applications. Finally, a case study which involved significant collaboration with a well-known safety-related solution provider in the nuclear industry is described, where Coriolis flowmeters were required to provide accurate measurement for the boric acid make-up subsystem in a pressurised water reactor system. Design checks and special considerations for the fabrication and testing according to the rules of ASME Code Section III together with a quality assurance procedure as described in the case study showed that the developed Coriolis flowmeter can meet the specific requirements of nuclear applications in terms of both measurement performance and safety concerns.
Reactor Coolant System Branch Line Pipe Break Analysis for Shin-Kori Units 3 and 4
Authors: Byun Young Jin , Sul Young Sil , Lee Kwang Won , Seo Jong Tae
Published: April 2011

Shin-Kori Units 3&4 (SKN 3&4) are the first APR1400 (Advanced Power Reactor 1400 MWe) nuclear power plants being under construction in Korea. The capacity and size of main components in the APR1400 nuclear steam supply system have been increased from the OPR1000 (Optimized Power Reactor 1000 MWe) which is the reference plant in operation. Especially, the increased pipe size of the main steam and the economizer feedwater lines cause more severe excitations such as the nozzle thrust, jet impingement, and nozzle reactions when those lines are postulated to break at terminal ends or intermediate locations. In this paper, the major mechanical effects on the Reactor Coolant System (RCS) due to the branch line pipe breaks for SKN 3&4 are investigated; the support loads of the reactor vessel, steam generators, reactor coolant pumps are evaluated. In this investigation, the RCS is modeled using 3-dimensional lumped mass and beam elements, and the nonlinear time-history structural analyses are performed. The analysis results show that the RCS dynamic responses are governed by the main steam line breaks and feedwater economizer line breaks.
Heat-Loss Calculations in a SCWR Fuel-Channel
Authors: Peiman Wargha , Saltanov Eugene , Gabriel Kamiel , Pioro Igor
Published: April 2011

The objective of this paper is to calculate heat losses from a CANDU-6 fuel-channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from fuel-channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the fuel-channel were calculated along the heated length of the fuel-channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The fuel-channel was divided into small segments and for each segment thermal resistances of the fuel-channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per fuel-channel when the pressure of the annulus gas was 0.3 MPa.
Development and Application of a Risk Monitor for Nuclear Power Plant
Authors: Wu Yi-can , Li Ya-zhou , Hu Li-qin , Luo Yue-tong , Wng Jin , Gu Xiao-hui , Wang Jia-qun , Wang Fang , Chen Tao , Qi Ji
Published: April 2011

Risk monitor, which has been widely used in the progress of RID (Risk Informed Decision) in a NPP (Nuclear Power Plant), is a plant specific real-time analysis tool to determine the instantaneous risk based on actual plant configuration. Based on wide investigation of challenges and technical issues during the development of a risk monitor, a prototype named Risk Angel has been designed by FDS Team in collaboration with several institutes and universities. An overview of the architecture and main functions of Risk Angel were introduced in this paper, as well as the quantitative approach, the calculating engine and the model development. Risk Angel has been applied in a nuclear power plant in P.R. China.
Numerical Analysis for the Swirl Tube Structure of the Cooling System in Neutral Beam Injection System
Authors: Tao Ling , Xie Yuanlai , Hu Chundong
Published: April 2011

Temperature control of High-Heat-Flux (HHF) components in Neutral Beam Injection (NBI) system is a very important issue for realizing the high-parameters steady state condition of NBI system. As one of the main HHF components, the V-shaped Residual Ion Dump (RID) with cross-sections structure is studied in this article. In order to enhance heat transfer of the fluent, the initial design with the swirl tubes is presented, and one side of the symmetrical ion dump under the designed inlet pressure is modeled. The flow field inside swirl tubes and the temperature distribution of the upper surface of the dump are simulated. This study provides a sample of solving the problem about how to protect all of the HHF components.
Study of the Characteristic Response of Pressure Control System in Order to Obtain the Design Parameters of the New Control System MARK VI Turbine in Cofrentes Nuclear Power Plant
Authors: Bucho´ Laura , Palomo Mari´a Jose´ , Vaquer Juan Ignacio , Lo´pez Bele´n , Rui´z Gregorio , Mora Jose´ Antonio , Verdu´ Gumersindo
Published: April 2011

This paper presents the results obtained from the IBE-CNC/DAQ-090827 project, conducted by the company “Titania Servicios Tecnolo´gicos, S.L.” in collaboration with the “Instituto de Seguridad Industrial, Radiofi´sica y Medioambiental” (ISIRYM), in the “Universidad Polite´cnica de Valencia”, for the company “Iberdrola Generacio´n S.A”. The objective is the acquisition of the pressure sensor signal and the measurement at points C85 and N32 from the cabin of the Turbine Control System in Cofrentes Nuclear Power Plant. With the study of previous data, one can obtain the Bode plot of the crossed signals as requested in the technical specification IM 0191 I. Frequency response (i.e. how the system varies its gain and offset depending on the frequency) defines the dynamics.
Stratified Flows in Pipes: A Parametric Study Towards Fatigue Relevance
Authors: Cizelj Leon , Simonovski Igor
Published: April 2011

Stratified flows may form in pipelines under certain conditions and could lead to increased fatigue loading that was only marginally accounted for during the design phase of the second generation of nuclear power plants designed in accordance with the ASME Boiler and Pressure Vessel Code. Extension of operational license would require explicit account for fatigue loads imposed by stratified flows. This typically involves rather complex state-of-the-art computational technology, which may in some cases be combined with measurements of the temperatures at the outside surfaces of pipes, which comprise pressure boundary of the reactor coolant. A parametric study using detailed finite element analysis of the entire span of the pipe has been performed to quantify the possible range of fatigue loads and fatigue usage factors. The example taken was a typical pressurized water reactor pressurizer surge line containing stratified flow of cold and hot water. The investigated parameters include the film coefficients governing the heat transfer from the both fluids to the pipe wall and the velocity of the interface between the cold and hot water. The main results include the expected ranges of fatigue loading given the range of investigated parameters. It is clearly shown that the choice of the film coefficients is essential to arrive at reliable fatigue estimate. Additionally, predictions of readings provided by hypothetical thermocouples at the pipe outer surface are provided. Some of their limitations are identified and discussed.
An Accurate Probabilistic Model for Estimating the Life Cycle Cost of Degrading Components in Nuclear Power Plants
Authors: Cheng T. , Pandey M. D. , Xie W. C.
Published: April 2011

Degradation of systems and components operating in harsh environment has an adverse effect on safety and reliability of nuclear power plants. Condition-based maintenance (CBM) programs are used to preventively maintain degrading components, which minimize the risk of failure. However, maintenance programs can be costly due to frequent inspection, increased outage time, and redundant maintenance of functional components. The optimization of maintenance programs over the life cycle of systems is an important issue for the plant managers. The paper presents an advanced model for the evaluation of life cycle cost of degrading components in the nuclear plants, which can be used for the maintenance optimization. The proposed model is based on the more precise finite time horizon formulation, instead of using asymptotic formulae that may lead to inaccurate results in practical settings.
Methods to Reduce CIPS/CILC Risk for the Zero Fuel Failure by 2010 Initiative
Authors: Wang Guoqiang , Young Michael Y. , Byers William A. , Krammen Michael A.
Published: April 2011

In order to support the industry goal of zero fuel failures by 2010 and beyond, established by the Institute of Nuclear Power Operations (INPO), this paper describes how the Crud Induced Power Shift (CIPS) and Crud Induced Localized Corrosion (CILC) risk is being addressed by Westinghouse Electric Company LLC. As one of the key fuel failure mechanisms, CIPS/CILC risk assessments are being targeted, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions will be discussed. Plant and experimental data will be presented to show the current state of CIPS/CILC risk assessments for Westinghouse fuel. Finally, suggestions and other initiatives that specifically target reducing CIPS/CILC risk will be described for Pressurized Water Reactors (PWRs) supported by test and/or analysis results.