bluuebunny/crossref_metadata_2025 huggingface dataset:
Meshless Local Petrov-Galerkin Method for the Steady Magnetohydrodynamic (MHD) Flow in Rectangular Ducts With Arbitrary Wall Conductivity
Authors: Cai Xinghui , Su Guanghui , Qiu Suizheng
Published: April 2011

In this paper, a meshless local Petrov-Galerkin (MLPG) method is given to obtain the numerical solution of the coupled equations in velocity and magnetic field for the steady magnetohydrodynamic (MHD) flow through a straight duct of rectangular section. Mehdi Dehghan has applied MLPG method to solve the MHD flow control equations at Hartmman numbers less than 40. The integration subdomain is properly adjusted to carry out the computations for Hartmman numbers from 5 to 400. Numerical results show that the MLPG method with adjusted integration subdomains can compute MHD problems not only at low values but also at moderate values of the Hartmann number with good accuracy and convergence.
On Fatigue Verification of Class 1 Nuclear Power Piping According to ASME NB-3600
Authors: Zeng Lingfu , Jansson Lennart G. , Dahlstro¨m Lars
Published: April 2011

In this paper, we address fatigue verification of Class 1 nuclear power piping according to ASME Boiler & Pressure Vessel Code Section III (ASME III), NB-3600, and several relevant issues that are often discussed in connection to the power uprate of several Swedish BWR reactors in recent years. We review first the basic requirements and their verifications using finite element analysis in detail. Thereafter, we clarify a so-called simplified elastic-plastic discontinuity analysis for further verification if the basic requirements found unsatisfactory, and examine necessary computational procedures for evaluating alternating stress intensities and cumulative damage factors. Our emphasis is placed on alternative verification procedures, which do not violate the general design principles upon which ASME III NB-3600 is built, when fatigue damage usages predicted by the simplified elastic-plastic discontinuity analysis are unaccepted. An alternative which employs a non-linear finite element computation and a refined numerical approach for re-evaluating the cumulative damage factors is suggested. Concluding remarks are given.
Pressure Drop of Perforated Orifice in Single-Phase Flow
Authors: Huang Shan-Fang , Ma Tai-Yi , Yexin Ou-wen , Yang Yan-hua , Wang Dong , Yan Xiao
Published: April 2011

This study studied the performance of perforated orifices by a new method, namely field synergy theory (FST). Pressure loss was obtained theoretically by connecting flow field and pressure loss, and pointed out the direction to optimization. To validate the theory, pressure loss was measured experimentally under different orifice structures by changing the parameters of porosity, thickness, hole diameter, hole distribution, and etc. Theoretical results showed that pressure loss was determined by the synergy of velocity and velocity gradient over the flow field passing the orifice. Pressure loss increased as the synergy being stronger. To decrease pressure loss, it was the most effective to decrease the synergy with the biggest weight. The comprehensive performance of a perforated orifice is better than a standard one. All the structure parameters should be optimized to improve the performance of a perforated orifice.
Iconic Representation Extension of Event Sequence Diagram Framework for Nuclear Power Plant
Authors: Xie Haiyan , Cai Qi , Zhang Yangwei
Published: April 2011

As a dynamic reliability methodology, current Event Sequence Diagram (ESD) framework becomes an important tool to model the dynamic situations. It has been employed in the risk analysis in nuclear and space industries. It has been proved that ESD can be helpful for the operators of nuclear power plant to understand the accident scenarios and monitor progress during accident evolution. Despite this, there are some limits for the ESD framework in the practical application, such as the icons in the ESD framework are inadequate, so many dynamic situations are difficulty or complicated to describe. As for nuclear power plant, this paper proposes to extend its current iconic system to simpler and better model the dynamic process. Based on current ESD icons, we have created many other icons to represent the dynamic situations of the system and the component according to the characteristics of nuclear power plant. Some complex situations and interactions in the operation and accident are expressed by only one or two icons, so the system dynamics and accident evolution can be clearly described and simply translated into ESD models. Then the loss-of-coolant accident is used as the example to show the application of extended icons. These new icons greatly reduce the size of ESD model, which is well for its building and quantitative analysis. Most of the extensional ESD icons are also suitable for complex systems in other research fields and the ESD methodology may become popular for dynamic reliability analysis in the industry application.
Study of Unique Foam Detecting Sensor
Authors: Li Shengqiang , Jiang Shengyao , Zhong Shuoping , Wang Wenran , Bo Hanliang
Published: April 2011

Liquid level measurements have been studied for a long time. Many research systems have been constructed. Some sophisticated systems have successfully experiences in industrial applications for years. But, in some specific situations, the liquid level is replaced by multiphase media level, such as bubbly flow level or foam level. Those have been widely observed in waste disposal process, chemical engineering fields and even during process of Lost Of Coolant Accident (LOCA) events in the Pressure Water Reactor (PWR). As the foam is a special phase compared with liquid, gas or steam, the unique mechanics, thermodynamics and optics characteristics make the measurement for it much difficult. It is still a Cutting-edge topic in level measurement field by now. A unique foam level sensor is described in this paper. The structure, principle are analysis in detail.
The Overlapping Peaks Decomposition Method of Nuclear Spectrum Based on the Statistical Clustering Model
Authors: Huang Hongquan , Fang Fang , He Zishu , Wang Min , Xu Changming
Published: April 2011

The full energy peaks always overlap each other when carrying out radioactive spectrum measurement. In this paper, combining GMM statistical clustering model with EM iterative algorithm, the overlapping peaks of nuclear spectrum with no background has been decomposed. No matter how many overlapping peaks we can use this decomposition method, and ensure that the statistical errors close to zero. It has great practical value for radionuclide with a quantitative or qualitative analysis.
Introduction of Interim Storage of SNF in INET
Authors: Huang Peng , Chen Xiaoming , Jia Lixin
Published: April 2011

There are three research reactors of different type in Institute of Nuclear and New Energy Technology of Tsinghua University (INET), they are Swimming Pool Shielding Reactor (SPSR), 5MW Nuclear Heating Reactor (NHR-5) and 10MW High Temperature Gas Cooled Reactor (HTR-10). The interim storage facilities for Spent Nuclear Fuel (SNF) are applied in these reactors except NHR-5, the SPSR adopts the wet storage for its SNFs, HTR-10 accepts the dry storage due to the ceramic structure of its SNFs. The practical storage conditions including SNFs features, space distribution, SNFs transportation, shielding measures and the safety analysis results including radioactive activity, shielding effect, the total amount of radioactive product leakage to environment are demonstrated in this paper.
Modeling and Study of Piezoelectric Power Generators
Authors: Tan Xiaohui , Soh A. K.
Published: April 2011

In the present study, a type of power generators of cantilever beam structure has been studied. A dynamic model describing the power generator, which takes into consideration the vibration of the mechanical structure, the power harvesting circuit, and the coupling between them, is proposed. A set of partial differential equations describing the dynamic model are obtained. Stability analysis is then carried out to analyze elementary properties of the generator. By using modal decomposition and state space representation, the set of partial differential equations could be easily solved no matter what the external excitation is, harmonic or non-harmonic, transient or continuous, at or off resonant frequency. The characteristics of the generators are analyzed and some recommendations on design of efficient power generators are provided.
Kinematics Analysis of Double Seal Door for ITER Remote Handling Transfer Cask System
Authors: Qin Shijun , Song Yuntao , Li Ge , Yao Damao
Published: April 2011

The moving direction of double seal door (DSD) of ITER remote handling transfer cask and the force of hydraulic pole will change significantly at the guide rail inflexion position (GRIP) which is a mutant site, so it is very possible to make the structure damage or the system failure at the GRIP. In this paper, the kinematics simulation and analysis of DSD were done based on special constitution restriction and working process by software ADAMS. The stress distribution of guide rail and hydraulic pole were obtained by the above simulation, at the same time the optimal GRIP was confirmed according to the force analysis result. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of kinematics analysis for other important components of ITER.
Preliminary Discussion on Comparison and Analysis on Nuclear Power Plant Materials Standards of ASME and RCC-M
Authors: Ning Dong , Wang Yongdong , Li Hui , Bao Zhanggen
Published: April 2011

This paper is to compare the sections on materials between ASME-BVC and RCC-M code. This paper states the layout structure and the content about NPP materials specified in the two codes and shows the comparison results. This paper emphasizes particularly on the basic differences between two codes; ASME code focuses on fracture toughness requirements for pressure-retaining ferrite materials and RCC-M code focuses on Susceptibility to intergranular corrosion for austenitic and austenitic-ferrite stainless steel.
Commercial-Grade Item Dedication for Nuclear Safety-Related Application
Authors: Hsu Yao-Tung , Chen Ming-Huei
Published: April 2011

This paper describes the general practice and lessons learned from Commercial-Grade Item dedication for nuclear safety-related system applications in Taiwan. The dedication process qualified the commercial off-the-shelf components to be applied as basic components. In past fifteen years, Institute of Nuclear Energy Research (INER) has actively performed the dedication service to help local nuclear power plants solve their procurement problems of nuclear grade items, due to reduced availability of qualified suppliers and/or obsolete issues of qualified components. The Scope of dedication includes material, electrical and mechanical components located in mild and harsh environment. Thousands of components such as piping, fitting, breaker, relay, motor, and control device etc., have already been dedicated to and successfully used in local nuclear power plants. The Commercial-Grade Item dedication process is based on EPRI documentations. Besides, the technical evaluation and equipment qualification are included during the dedication process. The requirements for equipment qualification are described in IEEE standards. Although the codes and standards for dedication in Taiwan refer to those in USA, the challenges may happen due to different regulators, utility, manufacture’s quality culture, and personal responsibility. The key to successful dedication will be dependent on the well-defined component requirements and good project planning. This paper introduces the self-reliant experiences in dedication and economic benefit to local nuclear power plants.
Assessment for Effect of RCS Depressurization and Reactor Cavity Flooding Using PSA Approach
Authors: Peng Changhong , Zhang Ning , Liu Pingping
Published: April 2011

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will reduce the amount of radionuclides released to the environment during the accident. The plant studied is a three-loop pressurized water reactor (PWR) with large dry containment. The RCS depressurization and reactor cavity flooding can be used as an accident management strategy. Then, the decrease of LERF (Large and Early Release Frequency) is quantified using PSA approach. It can be found that strategy of RCS depressurization and reactor cavity flooding can mitigate the result of severe accident effectively.
Prediction of Radioactive Corrosion Product Transfer in Primary Systems of Japanese Prototype Fast Breeder Reactor MONJU
Authors: Matuo Youichirou , Hasegawa Masanori , Maegawa Yoshiharu , Miyahara Shinya
Published: April 2011

Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance with no breached fuel in FBR plants. CP is produced in the core region by activation of fuel cladding and sub-assembly wrappers, and they are transported to the primary circuit with sodium flow and deposited on the wall of the primary piping and components. In order to establish the techniques of radiation dose reduction for of personnel, program system for corrosion hazard evaluation code “PSYCHE” has been was developed. The PSYCHE code is based on the solution-precipitation model. The density of each deposited CP and dose rate of primary coolant system in MONJU was estimated by using the PSYCHE and QAD-CG code. From the calculation for MONJU, it was predicted that deposition inventory of 54Mn would be about ten times more than that of 60Co in 20 years. In particularly deposition density; 54Mn would deposit mainly in the primary pump and cold-leg. On the contrary, 60Co depositions to not only the primary pump but also the Hot-leg. In the result, the future values of dose rate around primary piping system were estimated to be saturated at 2∼3 mSv/h. The calculation result of CP suggests that the analysis code and model make it possible to differentiate behavior of these nuclear species, definitely. Using the analysis code, the prospects for the future operations of MONJU could be established. Moreover, it is necessary to examination of improvement that the innovative calculation procedure and physics model, for applied to next generation LMFBR such as JSFR.
Nuclear Coal Gasification for Hydrogen and Synthetic Fuels Production
Authors: Verfondern Karl , von Lensa Werner
Published: April 2011

The most abundantly available fossil fuel on Earth is coal. For countries like China, the USA, South Africa, or Germany, coal plays a dominant role as energy resource. The introduction of nuclear energy into coal refinement processes would be a significant contribution to the saving of resources, lowering specific carbon emissions and reducing dependencies on oil and natural gas imports. In Germany, comprehensive R&D activities were conducted within the project “Prototype Plant Nuclear Process Heat” (PNP) to investigate the utilization of nuclear energy from a pebble-bed HTGR in both steam-coal gasification and hydro-gasification. A major component to be newly developed was the gas generator. Its operation on semitechnical scale confirmed the feasibility of allothermal, continuous coal gasification under nuclear conditions. A key problem remained the selection of appropriate high temperature materials for gas generator and other high temperature heat exchanging components. The project was accompanied by comprehensive safety studies targeting tritium contamination and consequences of potential explosions of flammable gas mixtures. Future activities could take benefit from a reevaluation of the studies conducted in the past by comparing HTGR process heat applications against current technologies. Fossil fuel market conditions and environmental effects shall be considered. Superior safety features and high reliability are prerequisites for the introduction of nuclear process heat and nuclear combined heat and power.
Validation of SIMMER-III Neutronics Module for the Simulation of Reactivity Injection Accident in Material Testing Reactors
Authors: Chegrani Yacine , d’Aletto Corinne , Di Salvo Jacques , Ivanov Evgeny
Published: April 2011

The “Institut de Radioprotection et de Suˆrete´ Nucle´aire”, as the technical support of the French Safety Authority, carries out studies and research to analyze and assess the safety of all nuclear plants. In this frame IRSN studies the feasibility of modeling Material Testing Reactor core with SIMMER-III code, for simulation of reactivity initiated accidental transients. The SIMMER-III multi-physics code system was initially developed for mechanistic safety analyses of liquid metal cooled fast reactors while employing coupled spatial neutron kinetics and thermal hydraulics models. Neutronics and thermal-hydraulics SIMMER-III models have been extended to safety analyses for water cooled and moderated reactors. The use of a code like SIMMER-III requires approximations; it computes a simplified R-Z geometry and chemistry description of the core that must be validated. The methods applied consist here in developing models of the same reactor on several scales of detail. The first step is the validation of the cross section condensation for deterministic APOLLO2 calculation against Monte Carlo TRIPOLI4 2D model. Temperature effects, kinetic parameters and void coefficients on the whole core are then calculated on a 2D APOLLO2 model, using the Method of Characteristics. These parameters are also computed with a 3D combined transport and diffusion calculations by means of APOLLO2/CRONOS2 calculations, validated against a TRIPOLI4 3D precise reference model. The final step is the validation of the simmer-like R-Z geometry in APOLLO2 Sn and Pij. Finally, an R-Z geometry has been computed in SIMMER-III, for the calculation of the kinetic parameters and temperature coefficients. This validation method has been applied to Jules Horowitz Reactor, a French Material Testing Reactor currently in commissioning by the CEA. This leads to conclude that discrepancies due to simplifications are acceptable. Moreover SIMMER-III shows quite a good agreement with CEA ring calculation on the kinetic parameters. Concerning neutronics feedbacks coefficients, further analyses remain necessary.
Development of High Temperature Gas-Cooled Reactor Fuel for Extended Burnup
Authors: Ueta Shohei , Aihara Jun , Honda Masaki , Furihata Noboru , Sawa Kazuhiro
Published: April 2011

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].
Flow Control in a Mixing-Vane Grid to Enhance Thermal Hydraulic Performance
Authors: Ikeno Tsutomu , Sasakawa Tatsuya , Kakinoki Shumpei , Murase Momonori
Published: April 2011

To investigate the effect of mixing-vane shape, heat flux at departure from nucleate boiling (DNB) and pressure loss were measured. Computational fluid dynamics (CFD) was utilized to discuss the flow control. The pressure loss and the DNB tests were performed in a water and a Freon loops, respectively. Two mixing-vanes were designed to have same projection area but different inclination. The rod-bundle was 5 by 5 and 17 by 17 respectively at the water and Freon tests. The experimental results showed that the slightly inclined mixing-vane produced the same DNB heat flux as the deeply inclined mixing-vane and did smaller pressure loss than it. Pressure loss of the two mixing-vane grids was different in spite of the same projection area. The result of CFD showed a swirl flow decaying along the main stream in the axial direction. The swirl was stronger in the deeply inclined mixing-vane, however it decayed faster whereas one maintained long in the slightly inclined mixing-vane. This result suggested that the deep inclination caused a steep change in axial momentum to induce strong turbulence diffusion. This flow structure did not change the DNB heat flux because the two-phase discontinuity dominated the phenomena. This study provided a successful example of flow control in a mixing-vane grid.
Threshold Stress Intensity Factor for Delayed Hydride Cracking of a Recrystallized N18 Alloy Plate Along the Rolling Direction
Authors: Sun Chao , Ying Shi H. , Tan Jun , Peng Qian , Dai Xun
Published: April 2011

The objective of this study is to obtain the threshold stress intensity factor, KIH, for an initiation of delayed hydride cracking (DHC) in a recrystallized N18 (Zr-1.05Sn-0.27Nb) alloy plate which was manufactured in China, gaseously charged with 60ppm of hydrogen by weight. By using both the load increasing method (LIM) and load drop method (LDM), the KIH’s along the rolling direction were investigated over a temperature range of 150–255°C. The results showed that KIH along the rolling direction was found to be higher in the LIM than that in the LDM. In the LIM, KIH’s of the N18 alloy plate appeared to be in the range of 31–32.5 Mpa·m1/2, and KIH in the LDM appeared to be in the range of 27.5–28.6 Mpa·m1/2. This means that the N18 alloy plate has high tolerance for DHC initiation along the rolling direction. The texture of a N18 alloy plate was investigated using an X-ray diffraction (XRD) and the KIH was discussed based on texture and analytically as a function of the tilting angle of hydride habit planes to the cracking plane.
Evaluation Technique With MCNP for Temperature Dependent Problems
Authors: Fujime Hiroyuki , Abe Shinji , Yamaji Kazuya , Sato Daisuke , Matsumoto Hideki
Published: April 2011

Monte Carlo calculation has come to be used as reference solutions instead of experiments in nuclear design code validation and verification (V&V), although comparisons with measurements are still indispensable for V&V in nuclear design. MCNP [1] is one of the most famous Monte Carlo codes widely used in the world. Many reference results are given for the analyses of critical experiments. When using the use MCNP calculations for validations of commercial design codes, we will face to a problem of lacking temperature dependent cross-sections. The cross-sections can be generated by the NJOY code [2]. However, if the model has complex temperature distribution, many NJOY calculations are necessary. Besides, if the temperature profile changes with fuel power and so on, many NJOY calculations have to be performed again and again. These back and forth procedures make us give up using MCNP for commercial LWR calculations. In order to solve this problem, we propose an easy approximation to solve the temperature problems using MCNP. Note that our technique does not require any code modifications.
Dynamic Simulation of Marine Nuclear Pressurizer
Authors: Chen Zhiyun , Chen Wenzhen , Luo Lei
Published: April 2011

According to the specific structure of a marine reactor pressurizer and its operating characteristics of frequent pressure and water level fluctuation, a new non-equilibrium four region mathematical and physical model is established for the first time to fulfill the aim of fast and precise simulation. The model includes the bubble rising, condensate dropping, wall condensation and heat transfer through outside wall while the heat and mass transfer at the liquid and gas interface is neglected. The initializing program is also introduced. The Simulink is adopted as the simulating tool and the corresponding Simulink module is established. The simulation proves to be very fast in a common microcomputer, which indicates that Simulink is efficient and is suitable for the fast simulation of marine reactor with the characteristics of frequent variations of power and corresponding physical and thermal parameters. A typical process with surge flow varying very acutely and complicatedly is adopted as an example. The variation of the pressure and the water level of the process by the new model agree well with the results of Relap5/Mod3.3, which proves the validity of the model and some simplification adopted. As one of the main parts of the fast simulation program under development for in the operating field use for a marine nuclear power plant, the achievement of the fast simulation of the pressurizer is of significant importance.
Thermal Conduction Experiments on Compacted GMZ Bentonite Blocks as the Potential Buffer Materials for HLW Disposal Purpose
Authors: Ren Guang-yuan , Liu Xiao-dong , Chen Qing-chun , Zhu Guo-ping , Chen Quan-shui
Published: April 2011

Compacted Na-bentonite blocks, with the original water content of 9% and compacting density of 1800kg/m3, from the Gaomiaozi (GMZ) deposit in Inner Mongolia Autonomous Region have been experienced about two months experiment in order to reveal the transmission patterns of temperature and the heat induced moisture movement in GMZ Na-bentonite blocks under high-level radioactive waste repository-like conditions. Based on the design, temperature and humidity composed micro-sensors have been used and allocated in different positions in the bentonite blocks as the check points of the experiment. There are two thermal conduction stages in GMZ bentonite blocks in a closed cylinder heat conduction system. In the first stage, the temperature at the check points in the bentonite blocks near the heater in the centre of the cylinder increased to 55°C in 20 hours, and the temperature at the check points distributed near the edge of the cylinder reached about 48°C in the same period. The temperature was maintained at about 60°C for about 35 days in the first stage. The heat induced moisture movement in bentonite blocks in the cylinder is different from changes of the temperature. The relative humidity at the check points near the heater increased quickly at the beginning of the heating, and then slowly decreased with the temperature maintained at 60°C. The average radial temperature gradient (GT) and radial relative humidity gradient (GHR) in bentonite blocks are 0.85°C/cm and 1.32%/cm respectively in the first stage. At the end of the first thermal conduction stage, the temperature of heater was slowly increased into 85°C and maintained this temperature for about 25 days as the second thermal conduction stage. The transmission patterns of temperature and the heat induced moisture movement in bentonite blocks are similar to the first stage. However, the radial temperature and humidity gradients are higher than that in the first stage. The average radial temperature gradient (GT) and radial relative humidity gradient (GHR) in bentonite blocks are 1.57°C/cm and 1.89%/cm respectively in the second stage.
Thermal Hydraulic Analysis Towards a Robust Design of Leak Collection Tray for Pool Type Sodium Cooled Fast Reactors
Authors: Sharma Anil Kumar , Velusamy K. , Kasinathan N. , Chellapandi P. , Chetal S. C. , Raj Baldev
Published: April 2011

To protect the sodium cooled FBR plant against the hazardous effects of sodium leak into the ambient, one of the passive protection devices used is the Leak Collection Trays (LCT) below the secondary sodium carrying pipelines in the Steam Generator Building (SGB). The design of LCT is based on immediate channeling of burning liquid sodium on the funnel shaped ‘sloping cover tray’ to the bottom ‘sodium hold-up vessel’ in which self-extinction of the fire occurs due to oxygen starvation. In the secondary heat transfer circuits of FBRs, leakage of liquid sodium from the pipelines is postulated as one of the design basis accidents with probability of occurrence at 10−2 per reactor year. LCT collect the leaked sodium in a hold up vessel, suppress the sodium fire due to oxygen starvation and guide the sodium to an inerted ‘sodium transfer tank’ located at the bottom most elevation of the SGB. The procedure of draining the leaked sodium into the transfer tank has been envisaged as a defense in depth measure against the handling of un-burnt sodium and to guard against larger leak rates than that can be handled by the LCT effectively. Towards this, a network of carbon steel pipelines are laid out connecting all the LCT and the transfer tank through headers in strategic locations, each having a fusible plug. The fusible plug separates the air environment in LCT and argon environment in sodium transfer tank. Woods metal is the preliminary choice for the fusible plug. It is an alloy of 50% Bi, 25% Pb, 12.5% Sn and 12.5% Cd with a melting point of 72°C. The transfer tank is filled with argon at ∼ 0.03 bars-g pressure. Both the header and the tank are at room temperature during normal conditions. Leaked sodium by virtue of its high temperature has to heat up the fusible plug to melt the same and drain into the transfer tank. Transient thermal hydraulic investigations have been carried out to predict the fusing characteristics of woods metal plug. The numerical results have been validated against analytical solutions for idealized conditions. Detailed parametric studies have been carried out with plug thickness as a parameter. It is established that effective melting of the plug and trouble free draining of the leaked sodium is possible for a 3 mm thick fusible plug.
Design of an Ethernet Bus Interface Controller in a Nuclear Power Plant Simulator
Authors: Lu Lixuan , Le Dong
Published: April 2011

Obsolescence presents great challenge to Nuclear Power Plants (NPP) and plant simulators around the world. Old designs will have to be either modified, or replaced by new designs, in order to simplify maintenance, increase availability and meet ever-increasing operational and training requirements. Control system upgrade and Distributed Control System (DCS) design for both old plants and new builds have become the center of interest. In a DCS, communication networks connect control systems together to allow the exchange of information and feedback. Among the many communication network protocols, Ethernet can be a promising one. This paper describes a new Ethernet Bus Interface Controller (eBIC) used in the Input/Output (I/O) system of a Nuclear Power Plant (NPP) simulator in Canada.
A Statistical Mechanical Equation of State for Fluid UF4 and ThF4
Authors: Wang Shisheng , Maschek Werner
Published: April 2011

The Song, Mason and Ihm [1–4] equation of state with modification of Tao and Mason [5], originally derived for spherical and molecular fluids, is applied for fluid uranium tetrafluoride and thorium tetrafluoride based on the available experimental data. The equation of state based on statistical mechanical perturbation theory with the perturbation scheme of Weeks, Chandler, and Andresen [6]. The prediction of constants applied in the equation of state is based on the work of Boushehri et al. [7–8] using heat of vaporization and liquid density at the triple point. The vapor correction term bases on the experimental data. The results show that this equation of state agrees reasonably well with the available experimental data.
Risk Importance Measures for Maintenance in Non-Coherent Systems
Authors: Tao Ye , Lu Lixuan
Published: April 2011

Probabilistic Safety Assessment (PSA) has been an important tool to assist Nuclear Power Plants (NPPs) management over the years. Through PSA, the weak links of the whole system can be identified by component importance measures. The importance measures can be classified according to risk significance and safety significance. They signify the role that each component plays in either causing or contributing to the occurrence of an undesired event. A wide range of importance measures have been developed over the years and most of them are geared towards coherent systems. Importance analysis of non-coherent system is rather limited. In this paper, a set of component importance measures for non-coherent systems is analyzed and investigated. The comparison and the selection of the most informative and appropriate measures for guiding the maintenance of a system are presented. A steam generator level control system of NPP is used to demonstrate the application of the results.
Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China
Authors: Wang Yan , Heng Xie
Published: April 2011

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.
Analysis of Thermal Stratification and Fatigue Stress for Pressurizer Surge Line
Authors: Yu Xiaofei , Zhang Yixiong
Published: April 2011

Thermal stratification of pressurizer surge line induced by the inside fluid brings on global bending moments, local thermal stresses, unexpected displacements and support loadings of the pipe system. In order to confirm the structural integrity of pressurizer surge line affected by thermal stratification, this paper theoretically establishes thermal stratified transient and studies the calculation method of thermal stratified stress. A costly three-dimensional computation is simplified into a combined 1D/2D technique. This technique uses a pipe cross-section for computation of local thermal stresses and represents the whole surge line with one-dimensional pipe elements. The 2D pipe cross-section model is used to compute elastic thermal stresses in plane strain condition. Symmetry allows half the cross-section to be considered. The one-dimensional pipe elements model gives the global bending moments including effects of usual thermal expansion and thermal stratification of each model nodes. This combined 1D/2D technique has been developed and implemented to analyze the thermal stratification and fatigue stress of pressurize surge line in this paper, using computer codes SYSTUS and ROCOCO. According to the mechanical analysis results of stratification, the maximum stress and cumulative usage factor are obtained. The stress and fatigue intensity of the surge line tallies with the correlative criterion.
Development of Ultrasonic Sensor for Under-Sodium Volumetric Inspection of Heat Exchanger Pipes
Authors: Takeishi M. , Yui M. , Kobayashi T. , Tsuri F. , Kotake S.
Published: April 2011

Under sodium sensor has been developed to inspect heat exchanger pipes in FBR plants. Several methods have been applied to manufacture the volumetric inspection sensor. Low temperature diffusion bonding method was selected to bond piezoelectric element and thin stainless steel plate together for the boundary between piezoelectric element and sodium. Under sodium test has been made to verify the detection ability of the newly developed volumetric inspection sensor. Under sodium tests obtained that the sensor could detect 20% depth EDM (Electro discharge method) slits in outer and inner surface of 1mm thickness pipe wall.
Conceptual Design of CANDU-SCWR With Thermal-Hydraulics Coupling
Authors: Yang Ping , Cao Liangzhi , Wu Hongchun , Wang Changhui
Published: April 2011

A CANDU-SCWR core is designed by using a 3D neutronics/thermal-hydraulic coupling method. In the fuel channel design, a typical 43-element fuel bundle is used, the coolant and the moderator are supercritical water and heavy water respectively. The thickness of the moderator is optimized to ensure the negative coolant coefficient during operation. With 1220 MW electric power, the reactor core is designed with a diameter of 4.8m and length of 4.95m, and there are totally 300 fuel channels, each of which consists of 10 fuel bundles. The inlet coolant temperature is set to be 350 °C °C and the operation pressure is 25 MPa. In order to flatten the radial power distribution, the loading pattern of the equilibrium cycle is optimized, and an optimized fuel management scheme is used with three batches refueling, burnable poison Dy2O3 is used to flatten the power peaking. The numerical results show that the average power density is 42.75 W/cm3, while the maximum linear element rate (LER) is 575W/cm. The average discharged burnup of the equilibrium is 48.3GWD/tU, and a high average outlet coolant temperature of 625 °C is achieved with a maximum cladding surface temperature less than 850 °C. Besides, the coolant temperature coefficient is negative throughout the cycle.
The Investigation of Classification Management for SSCS Susceptible to Ageing in PWR
Authors: Xue Fei , Liu Peng , Meng Xin-ming , Ti Wen-xin , Lin Lei , Wang Zhao-xi , Zhang Guo-dong , Jiang Jia-wang
Published: April 2011

A set of screening and classification method is proposed in this paper to manage the Systems, Structures and Components (SSCs) of the Pressurized Water Reactor (PWR) for the aging management. The method proposed and carried out in China Nuclear Power Plants (NPP) is based on the systematic aging evaluation method and the management method for the license renewal from International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC), whose successful management methods are discussed in this paper. Considering the importance of the classification management of the SSCs, the classification method is investigated from the aspects of the safety, reliability and replacement of the SSCs with the Nuclear Power Plant operation experience feed-back. With the classification management method, the SSCs are managed economically and effectively for the Nuclear Power Plants in China. The method also makes great foundation for the aging management and residual life evaluation of the PWR in China.
Study on Development of Nuclear Power Standard in China
Authors: Pan Jianjun
Published: April 2011

Nearly about 100 nuclear power national standards and 340 industry standards have been made in the field of nuclear power in China, which can partly meet the need of development of the second generation pressurized water reactor nuclear power plant. The situation of nuclear power fast development in China causes more standards to be made in the field of nuclear power, furthermore, the standard system and standardization have some problems now, to make more standards scientifically and solve existing problems, a number of measures should be taken in future.
Flow Distribution and Turbulent Heat Transfer Measurements in a Hexagonal LBE Rod Bundle
Authors: Litfin Karsten , Batta Abdalla , Class Andreas G. , Wetzel Thomas
Published: April 2011

In the framework of accelerator driven sub-critical reactor systems (ADS), heavy liquid metals (HLM), in particular lead or lead bismuth eutectic (LBE), are considered as coolant for the reactor core and the spallation target due to their efficient heat removal properties and high production rate of neutrons. LBE-flows are characterized by excellent heat conductivity and exhibit a low molecular Prandtl number of the order 10−2 leading to distinct thermal and viscous boundary layers and scale separation in both the time and spatial domain. Since the analogy of turbulent heat and momentum transfer is employed in common turbulence models but is not valid in HLM flows, commercially available fluid dynamic code systems cannot predict heat transfer adequately for such flows. In order to provide validation data and heat transfer correlations, a series of three major experiments has been launched at the KArlsruhe Liquid metal LAboratory (KALLA) of the Karlsruhe Institute of Technology and will be presented in this overview.
Study on the Predicting Model of Onset of Nucleate Boiling in Natural Circulation Based on Unascertained Mathematics
Authors: Zhang Ming , Zhou Tao , Liu Ping , Ran Ke , Sheng Cheng , Fu Tao , Xiao Zejun
Published: April 2011

Subcooled boiling in natural circulation has great influences on the flow and heat transfer characteristics of the medium in primary circuit loop of pressurized water reactor. Onset of nucleate boiling (ONB) is a key point in boiling heat transfer. According to analyze the existing models of ONB, the main factors that affect ONB are found out. The effects of main factors and the location of ONB are analyzed based on Yang Ruichang model in natural circulation, and found a model to calculate the location of ONB in natural circulation. A mathematical model is created using unascertained mathematics and compared with other models. The conclusion illustrates that the predicting model can describe the physical phenomena of subcooled boiling efficiently, and depicts the relation of certainty and uncertainty of influencing factors preferably by unascertained mathematics, which has particular methods to handle incomplete information, unsteady and inaccuracy numerical value, and variation quantity. The model reveals the real circumstance more accurately and plays an important role in security evaluation in nuclear power plant.
Three-Dimensional Seismic Analysis on Reactor Structure
Authors: Jiang Naibin , Zang Feng-gang , Zhang Li-min , Zhang Chuan-yong
Published: April 2011

The seismic analysis on reactor structure was performed with a new generation of finite element software. The amount of freedom degree of the model was more than twenty millions. The typical responses to operational basis earthquake excitation were given. They are larger than those with two-dimensional simplified finite element method, and the reasons of this phenomenon were analyzed. The feasibility of seismic analysis on large-scale three-dimensional finite element model under existing hardware condition was demonstrated, so some technological reserves for dynamic analysis on complicated equipments or systems in nuclear engineering are provided.
Computational Method for Crack-Opening-Area of Nuclear Pressure Pipe With Circumferential Through-Wall Crack
Authors: Liu Jun , Yan Bo , Jiang Naibin
Published: April 2011

The determination of crack-opening-area (COA) is very important in the application of LBB technique. A numerical method based on coupled finite element-meshless Galerkin method (FEM/EFG) is proposed to calculate the crack-opening-area of cracked pipes in nuclear power station. By means of the method by which the meshless nodes are used in the vicinity of crack front and the finite elements are applied in remained zone, the resolved domain is numerically discreted. This technique can speed up the set up of the numerical model and improve the computational precise. Based on the displacement field determined by the coupled FEM/EFG, a kind of surface integral method is used to accurately calculate the crack-opening-area. The correctness and validity of the presented method are demonstrated by comparing the COA results of centre-cracked panel under uniform load determined by the presented method with those from theoretic formula. Subsequently the numerical method for cracked panel is employed to calculate the crack-opening-area of nuclear pressure pipe. The numerical COA results are compared with those determined by engineering methods. It is shown that for thick-wall or middle-thick-wall pipes, the COA results obtained by the presented method are more conservative than those by the engineering methods, whereas for thin-wall pipes, the results by the later one are more conservative than those by the former.
Inspection of the Steam Generator Divider Plate
Authors: D’Annucci Filippo , Lecour Eric
Published: April 2011

The inspection of the steam generator divider plate becomes more and more important since indications have been detected in some steam generators in the area of the welds between tube sheet and divider plate. Remote operated manipulator and tools have been developed in order to minimize the radiation exposure of the personnel performing the inspections. Inspection techniques based on liquid penetrant, visual inspections have been developed and qualified to detect the indications. An ultrasonic inspection technique to determine the depth of the indications has been developed and qualified. The aim of this paper is to describe the inspection capabilities available to support the partition plate inspections.
Use PSA Methodology to Evaluate Baseline Risk Index for Initiating Events in Daya Bay NPP
Authors: Lu Yinjuan , Xi Haiying , Guo Jianbing
Published: April 2011

Baseline Risk Index for Initiating Events (BRIIE) is an integrated industry-level initiating event performance indicator that is risk informed. BRIIE is a performance indicator that provides a mechanism for determining the risk significance of changes in performance, at both the individual initiating event level and at the integrated cornerstone of safety level. This paper describes the process to use BRIIE to evaluate the initiating events in Daya Bay NPP. We established two indexes for BRIIE, one is based on NRC baseline, the other is based on the plant specific PSA initiating events frequency. Both indexes of BRIIE are calculated in two tiers.