bluuebunny/crossref_metadata_2025 huggingface dataset:
Muscle and Tendon Tissues: Constitutive Modeling and Computational Issues
Authors: Spyrou L. A. , Aravas N.
Published: April 2011

A three-dimensional constitutive model for muscle and tendon tissues is developed. Muscle and tendon are considered as composite materials that consist of fibers and the connective tissues and biofluids surrounding the fibers. The model is nonlinear, rate dependent, and anisotropic due to the presence of the fibers. Both the active and passive behaviors of the muscle are considered. The muscle fiber stress depends on the strain (length), strain-rate (velocity), and the activation level of the muscle, whereas the tendon fiber exhibits only passive behavior and the stress depends only on the strain. Multiple fiber directions are modeled via superposition. A methodology for the numerical implementation of the constitutive model in a general-purpose finite element program is developed. The current scheme is used for either static or dynamic analyses. The model is validated by studying the extension of a squid tentacle during a strike to catch prey. The behavior of parallel-fibered and pennate muscles, as well as the human semitendinosus muscle, is studied.
A Series Solution for the Diffraction of SH Waves by a Vertical Edge Crack
Authors: Tsaur Deng-How , Chang Kao-Hao , Hsu Ming-Sheng
Published: April 2011

Herein, a substitute series solution, which is unavailable in literature, pertaining to the antiplane diffraction problem by a vertical edge crack, is derived via the region-matching technique combining with the method of images. The dynamic stress intensity factor can be straightforwardly extracted from the expression of the radial stress field, fulfilling the crack-tip singularity inherently. Computed results of the extremely near-, near-, and far-field solutions agree well with those of exact analytical solution and available data. The proposed approach can be applied to cope with the diffraction problems by multiple cracks or branched cracks under SH-wave incidence.
The Implementation of Radiological Characterization for Reactor Decommissioning
Authors: Deng Junxian , Zhao Huasong , Li Xin , Deng Feng
Published: April 2011

The radiological characterization includes: the collection and review of historical file; the performing calculation of radionuclide inventory in the reactor; in situ measurement; sampling analyses; the review and evaluation of the data obtained; the comparison of calculated result with measured date etc. The special attention should put to the information from the key part in reactor for end state radiological characterization. The sampling from the “hot spot” should not be lost; the number of the sampling should reasonable base on reliable statistics. The radioactivity density for site release should comply with the guide, standard and regulation of international atomic energy agency and our country.
Selective Separation and Recovery of Cesium by Ammonium Tungstophosphate-Alginate Microcapsules
Authors: Mimura Hitoshi , Wu Yan , Wang Yufei , Niibori Yuichi , Yamagishi Isao , Ozawa Masaki , Ohnishi Takashi , Koyama Shinichi
Published: April 2011

A fine crystalline ammonium tungstophosphate (AWP) exchanger with high selectivity toward Cs+ was encapsulated in biopolymer matrices (calcium alginate, CaALG). The characterization of the AWP-CaALG microcapsule was examined using SEM/WDS, IR and DTA/TG analyses, and the selective separation and recovery of 137Cs were examined by the batch and column methods using simulated and real high-level liquid waste (HLLW). The free energy (ΔG0) of the ion exchange (NH4+ ↔ Cs+) for fine AWP crystals was determined at −13.2 kJ/mol, indicating the high selectivity of AWP towards Cs+. Spherical and elastic AWP-CaALG microcapsules (∼700 μm in diameter) were obtained and fine AWP crystals were uniformly immobilized in alginate matrices. Relatively large Kd values of Cs+ above 105 cm3/g were obtained in the presence of 10−3∼1 M Ca(NO3)2, resulting in a separation factor of Cs/Rb exceeding 102. The irradiated samples (60Co, 17.6 kGy) also exhibited large Kd values exceeding 105 cm3/g in the presence of 2.5 M HNO3. The Kd values in the presence of 0.1–9 M HNO3 for 67 elements were determined and the order of Kd value was Cs+ ≫ Rb+ > Ag+. The breakthrough curve of Cs+ had an S-shaped profile, and the breakpoint increased with decreasing flow rate; the breakpoint and breakthrough capacity at a flow rate of 0.35 cm3/min for the column (0.7 g AWP-CaALG) were estimated at 25.2 cm3 and 0.068 mmol/g, respectively. Good breakthrough and elution properties were retained even after thrice-repeated runs. The uptake (%) of Cs+ in simulated HLLW (28 metal components-1.92 M HNO3, SW-11, JAEA) was estimated at 97%, and the distribution of Cs+ and Zr/Ru into the AWP and alginate phases, respectively, were observed by WDS analysis. Further, the selective uptake of 137Cs exceeding 99% was confirmed by using real HLLW (FBR “JOYO”, JAEA). The AWP-CaALG microcapsules are thus effective for the selective separation and recovery of Cs+ from HLLWs.
Evaluation of Thermal Aging Embrittlement in Main Coolant Pipe Steel by Small Punch Test
Authors: Xue Fei , Yu Weiwei , Chen Xiao , Liu Yan , Shi Zhenhao , Shu Guogang , Wang Xitao
Published: April 2011

To evaluate the thermal aging embrittlement in the main coolant pipe material of nuclear power plants, a series of small punch testes at room temperature were performed on Z3CN20.09M duplex stainless steel aged at 400°C for up to 10000 hours, and then the fractured specimens were examined by the SEM. The test results indicate that the aging heat treatment led to a slight increase of the strength, and caused an initial increase on the small punch energy followed by continual decrease when the aging time extends 1000 hours. Besides, the fracture mode of specimens aged at 400°C for different time exhibited some discrepancy, but the difference is not clearly to reflect the thermal aging effect on the material.
Experimental Investigation of Copper/Stainless Steel Joints Formed by Vacuum Brazing
Authors: Ye Changqing
Published: April 2011

The article presents a study of two different brazing joints produced by dissimilar materials vacuum brazing. The junctions were obtained between copper or copper alloy and stainless steel. Different brazing parameters were used according to the different type of samples. By using optical microscope, scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS) and micro-hardness machine to analyze the microstructure of copper or copper alloy/stainless steel vacuum brazing joins. The test results showed that copper (T2)/stainless steel (1Cr18Ni9Ti) dissimilar materials were successfully bonded together by means of the advanced vacuum brazing technology (the grade of filler metal was B-Ag72Cu). The interface zone of copper (T2)/stainless steel (1Cr18Ni9Ti) brazing bonded joint included the copper side interface, the middle brazing transition zone and stainless steel side. Some defects such as microfissures were also found in the brazing seam between copper alloy and stainless steel composite components obtained by vacuum brazing using B-AgCu21Pd25 filler metal. They are mainly due to the process and geometry parameters, such as temperature and clearance.
Reliability Research of Power-Supply System in Nuclear Power Plant Based on Markov Theory and Bayes Network
Authors: Zhen Li , Sun Xinli , Ding Jian , Zhang Xia , Cai Xinghui
Published: April 2011

As components in an actual nuclear system have multi states and are dependent on each other, a method for system reliability analyse based on Markov theory and Bayes network is presented in this paper. The method resolves the problems of multi-state and dependence simultaneously, by utilizing Markov theory and Bayes network synthetically. Applying the method to analysing reliability of power-supply system in nuclear power plant, the system failure probability, RAW and posterior probability of each component are obtained, which are useful for system reliability evaluation, improvement of weakness and fault diagnosis. Synchronously, the method is proved feasible and effective for solving system reliability that contains dependent and multi-state components.
The Design of Incorporation Methods of Sheet Radioisotope Source Into Nuclear Battery and Its Encapsulation
Authors: Hu Rui , Wang Quanquan , Zhang Huaming
Published: April 2011

The problems of incorporating radioactive material into PIN-diode and those of encapsulating the nuclear battery were waiting for being solved in the phases of developing of beta voltaic effect nuclear cell. Aiming at the need of loading sheet radioisotope source Ni-63 at present, this paper designed its loading method on PIN-junction and encapsulation manner of nuclear batteries with energy transforming structures based on the photoelectric detector. Some parameters relating with encapsulation of prototype cell were advanced.
Equilibrium Partition Coefficients of Cesium and Iodine Between Sodium Pool and the Inert Cover Gas
Authors: Miyahara Shinya , Nishimura Masahiro , Nakagiri Toshio
Published: April 2011

Equilibrium partition coefficients were experimentally measured for volatile fission products of cesium and iodine between liquid sodium pool and the inert cover gas. In the experiments, the “transpiration method” was utilized in which the saturation vapor of sodium with cesium and iodine vapor in an isothermal evaporation pot was transported by inert carrier gas and trapped by filters outside the pot. The objectives of the experiments are to: a) Obtain the equilibrium partition coefficients of cesium and iodine at high temperature between 600 and 850 deg-C, and b) Study the dependence of the partition coefficients upon the concentration in the sodium pool. From the results of previous work and this study, the following empirical equations between the partition coefficients of cesium and iodine and the sodium pool temperature could be obtained: logKd(Cs)=2173/T−1.0487(from450to850deg-C)logKd(I)=−215/T−0.271(from450to850deg-C) These equations are consistent with Castleman’s theoretical equations. The partition coefficients of cesium measured at five different points of mole concentration in the pool were almost consistent with the theoretical values and decreased with the increase in the concentration. On the other hand, the measured partition coefficients of iodine increased with the increase in the concentration in the pool and this tendency was incompatible with the theoretical consideration. The reason of this discrepancy might be attributed to the formation of Na2I2 in the cover gas.
Application of Genetic Algorithm in Neutron Spectrum Unfolding
Authors: Wang Dong , He Bin , Zhang Quanhu
Published: April 2011

Bonner Spheres neutron spectrometer has been widely applied as neutron dosimeter, however the derivation of neutron energy spectrum from its measurement data is still a significantly difficult task. This unfolding problem is proved to be ill-posed, under-determined and have no exact solution. Two major require of the unfolding methods are accuracy and stability. Most unfolding methods try to search the solution that best fit the measurement data and the response function. As a universal optimization tool Genetic Algorithm shows its potential to solve this kind of problem. Through gene operation of every generation, GA could find the global optimal among the searching space. A new fitness function which contains a distance part and a penalty part was constructed in this research. The distance part is the square distance between the individual and the measurement data. The penalty part which is a function associated with the continuity of individual is used to avoid intensively change of unfolded data. Five classical neutron spectra were chosen as benchmark input spectra. The product of the benchmark spectra and the response function played as input measurement data of the unfolding program. The unfolded results showed good agreement with the real ones. The measurement data could be well reproduced by the unfolded results though the results had some difference with the real spectra.
Heat Transfer Experiments and Analyses on Double-Tube Bundle Heat Exchanger
Authors: Li Yonghui , Chen Minghui , Li Rizhu
Published: April 2011

Experiments were performed on the double-tube bundle heat exchanger. The hot side of the heat exchanger consists of parallel flow channels of the inner tubes and the space between the shell and the double-tube bundle, and the cold side is the narrow annular flow channels between the inner and outer tubes. The test fluid in both hot and cold sides is de-ionized water. The heat-transfer coefficients were obtained from the experiment in the range of Reynolds number from 2000 to 38000. The detached coefficients method was used in the absence of direct measurement of the tube temperature. It is found that the ranges of laminar flow in both internal and external flow passages are wider than regular circular tube. The heat transfer coefficient of internal side is lower than those from Dittus-Boelter formula, while the heat transfer coefficient of external side in this experiment is higher than that for the corrugated pipe. The heat transfer coefficient of the narrow annular channel is significantly determined by its structure.
An Investigation on the Minimum Inventory of Human System Interfaces for Plant’s Safe Operation
Authors: Yang Chih-Wei , Yenn Tzu-Chung , Cheng Tsung-Chieh , Chen Ming-Huei , Wu Tung-Ming
Published: April 2011

This paper investigates the minimum inventory (MI) of human system interfaces (HSIs) (i.e. alarms, controls, and displays) for plant’s safe operation and represents the analytic procedure on the MI of HSIs developed for the digital instrumentation and control (I&C) equipments in the main control room (MCR). The MI of HSIs in the MCR indicates the HSIs that the operator always needs available to: (1) monitor the status; (2) perform and confirm a reactor trip; (3) perform and confirm a controlled shutdown of the reactor; (4) actuate safety related systems; (5) analyze failure conditions of the normal HSIs; (6) implement the plant’s emergency operating procedures (EOPs); (7) bring the plant to a safe condition; (8) carry out those operator actions shown to be risk important by the probabilistic risk assessment (PRA). The proposed analytic procedure on the MI of HSIs in this study can be used to (1) identify the MI of HSIs and their design requirements; and (2) addresses design requirements and implementation for the MI of HSIs. The contribution of this study is to describe the MI of HSIs needed to implement the plant’s EOPs, to bring the plant to a safe condition, and to carry out those operator actions shown to be risk important by the PRA.
Multiphysics Modeling and Simulation for Stress Corrosion Cracking Considering Oxygen Atom Diffusion Along Grain Boundary
Authors: Igarashi Takahiro , Aoyagi Yoshiteru , Kaji Yoshiyuki
Published: April 2011

Aged degradations of structural materials of boiling water reactors by stress corrosion cracking (SCC) have been frequently reported. SCC is the results of the synergistic interaction of mechanical stress and corrosive environment, and the investigation of this phenomenon has been an important issue. Although many kinds of studies for SCC have been carried out, we have not clarified the fundamental mechanisms of SCC initiation and propagation yet. In the recent experimental studies, nano-scale observation around crack tips using transmission electron microscopy have shown three characteristics of SCC of nuclear structural materials as follows; the size of crack tip is nanometer order, the opening crack is filled with the oxides, and oxygen atoms exist in the grain boundary beyond the crack tips. The second and third ones show that the corrosive environment is mainly influenced on the SCC propagation behavior. Furthermore, electron back scatter diffraction pattern analyses have shown that about 10–20% of plastic strain exists around the crack tips and crack sides. The existence of oxygen atoms along grain boundaries and plastic strains around grain boundaries could be related to the crack propagation mechanism of SCC. In this study, in order to observe the influence of oxygen atoms on the SCC propagation behavior, the two-dimensional SCC propagation model considering diffusion of oxygen atoms along grain boundaries is developed. In this model, the stress distribution of polycrystalline system is obtained by the crystal plasticity theory, and the concentration of oxygen atoms depending on stress localization around cracks is calculated using the diffusion equation of oxygen atoms considering the stress gradient. The density distribution of oxygen atoms is adopted for the threshold of the crack propagation. Relation between the threshold of crack propagation as a viewpoint of density of oxygen atoms along grain boundaries and the geometry of SCC is discussed in this paper.
Dimensionless Analysis of Pressure Drop and Heat Transfer on HTGR Coupled With Closed Brayton Cycle
Authors: Yu Zhenjia , Yang Xiaoyong , Yu Xiaoli , Wang Jie
Published: April 2011

High temperature gas-cooled reactor with direct helium turbine cycle is based on the closed Brayton cycle. Its outstanding feature is the high efficiency of power generation. Pervious researches showed that recuperator was the key component to promote the cycle’s efficiency. And pressure drops in components were unavoidable in actual projects and had significant influence on cycle efficiency. A dimensionless model was proposed to analyze cycle’s features of HTGR coupled with gas turbine. The parameters’ effect on cycle’s efficiency was analyzed, with full consideration of the frictional and local pressure drops respectively. Under the restriction of materials and state-of-art of technologies, it showed that the cycle’s efficiency depended on compression ratio, recuperator’s effectiveness and pressure drops of components. However the pressure drop ratios of different components were inherently connected due to the closed cycle. Furthermore pressure drops inside the recuperator were also the function of effectiveness of the heat transfer based on the Reynolds analogy. Therefore cycle’s efficiency just depended on recuperator’s effectiveness with fixed compression ratio. So there existed optimal recuperator’s effectiveness and maximum cycle’s efficiency, which varied with the pressure ratio and other parameters as temperature ratio. The calculation also indicated that the pressure drop in pipes was close to that in heat exchangers. That was, the local pressure drop and frictional pressure drop should be considered respectively, and the local pressure drop made quite large reduction of cycle’s efficiency. The result also showed that local pressure drop had great influence on parameters such as optimal compression ratio and recuperator’s effectiveness.
Performance Test Results of the Supercritical CO2 Compressor for a New Gas Turbine Generating System
Authors: Aritomi Masanori , Ishizuka Takao , Muto Yasushi , Tsuzuki Nobuyoshi
Published: April 2011

Supercritical carbon dioxide (S-CO2) gas turbines can generate power at high cycle thermal efficiency, even at modest temperatures of 500–550°C, because of their markedly reduced compressor work near the critical point. Furthermore, the reaction between Na and CO2 is milder than that between H2O and Na. A more reliable and economically advantageous power generation system could be achieved by coupling with a sodium-cooled fast reactor. At Tokyo Institute of Technology, numerous development projects have been conducted for development of this system in cooperation with JAEA. Supercritical CO2 compressor performance test results are given as described herein. A centrifugal compressor is chosen for the performance test. Main compressor parts are stored in a pressure vessel. Maximum design conditions of the supercritical CO2 test apparatus are pressure of 11 MPa, temperature of 150°C, the flow rate of 6 kg/s and rotational speed of 24,000 rpm. The centrifugal compressor has an electric motor with permanent magnets on the rotor surface, with speed control by an inverter up to 24,000 rpm, a rotor shaft for the impeller, and a motor supported by gas bearings. Different compressor design points are examined using impellers of three kinds; test data are obtained using those impellers under steady state conditions with changing pressure, temperature, flow rate, and compressor rotor speed. The pressure ratio (compressor outlet pressure/inlet pressure) is obtained with the function of compressor rotational speed and the fluid flow rate. The data cover a broad region from sub-critical to supercritical pressure. Such data were obtained for the first time. No unstable phenomenon was observed in the area where the CO2 properties change sharply. Data of the pressure ratio vs. flow rate were coincident with the fundamental compressor theory.
Thermal Neutron Cross Sections of Cold-Moderators
Authors: Wang Guanbo , Zhong Bin , Wang Kan , Yu Ganglin
Published: April 2011

This paper reviews some moderators for cold neutron source. As a good cold moderator, solid methane was studied and evaluated using the new synthetic frequency spectrum theory (SFS). Due to its high proton density and easy handling, mesitylene (not included in ENDF/B) has also been considered as a very good moderator for cold neutron source. Evaluation of this material in different crystalline phases was done by using a preliminary frequency spectra built combining experimental and synthetic contributions. As a result, we have generated ACE format scattering data files by the NJOY code, with validations of comparing total cross sections from experiments.
Experience and Future Plan of Test Operation Using HTTR
Authors: Nishihara Tetsuo , Tochio Daisuke , Shinohara Masanori , Shimazaki Yosuke , Nojiri Naoki , Iyoku Tatsuo
Published: April 2011

The High Temperature Engineering Test Reactor (HTTR) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR is a graphite-moderated and helium gas-cooled reactor with thermal power of 30MW and the maximum reactor outlet coolant temperature of 950°C. Main objectives of the HTTR are to establish and develop HTGR technology and to demonstrate process heat application. The HTTR has conducted two test operations which are safety demonstration test and continuous operation. The safety demonstration tests focus on the verification of the inherent safety features of the HTGR that is the negative reactivity feedback effect of the core brings the reactor power safely to a safe and stable level without a reactor scram and the temperature transient of the reactor is slow in case of anticipated operational occurrences (AOOs). The safety demonstration tests include reactivity insertion test, coolant flow reduction test and loss of forced cooling (LOFC) test. Reactivity insertion test and coolant flow reduction test have been conducted since 2002. These tests demonstrated the inherent safety features of the HTGR in case of reactivity insertion and coolant flow reduction, and provided valuable data for code validation. LOFC test will start in the middle of 2010. LOFC is one of the important accident scenarios in the safety assessment of the HTGR. This test result will show extreme safety features of the HTGR and further improve the safety design approach of the HTGR. Obtained data can be useful to validate plant safety analysis codes. The continuous operation has been conducted to obtain plant data to establish HTGR technology and to demonstrate capability of the HTTR to supply stable heat to heat utilization system for long-term. Two operations of 30-day continuous operation in rated operation mode (in which designed reactor outlet coolant temperature of 850°C) and 50-days continuous operation in high temperature test operation mode (in which designed reactor outlet coolant temperature of 950°C) have been conducted so far. The 30-day continuous operation was achieved in 2007 and a good fuel performance to retain fission products within the coated fuel particle was clarified. The HTTR conducts 50-days continuous operation in 2010 to add useful operation data at high temperature to improve technical basis of HTGR and to realize high temperature heat application of HTGR.
Modelling of External Cooling for In-Vessel Corium Retention in VVER-440/213 Type Nuclear Power Plants
Authors: E´zso¨l Gyo¨rgy , Baranyai Ga´bor , Perneczky La´szlo´ , Szabados La´szlo´ , To´th Iva´n
Published: April 2011

Research and development programs of high safety significance have been going on for nuclear power plants of VVER-440/213 type to apply the in-vessel corium retention concept. The in-vessel retention (IVR) concept is based on the external reactor vessel cooling (ERVC) with the main objective to prove that the reactor pressure vessel (RPV) integrity can be preserved in accident sequences leading to core melt. One of the bases of programs was the SARNET project of European Commission, which focused on confirming the capability of the ASTEC code to simulate IVR, calculating thermal load caused by the corium. The ERVC concept is applied to the Paks nuclear power plant of VVER-440/213 type. For the experimental modelling of the ERVC, the CERES (Cooling Effectiveness on Reactor External Surface) facility was designed and constructed. The facility is a scaled down model of the cooling system intended to apply to the Paks NPP with 1:40 scaling ratio for the vessel external surface and 1:1 for the elevations giving the driving force for natural circulation. The heat load supplied to the model is provided by electric heaters. A large number of temperature, pressure, level, void and flow measurements are installed. A RELAP model of the CERES facility was developed and tested by pre-test results.
Determinant Methods for Solving Neutron Transport Equation in Unstructured Geometry
Authors: Wu Hongchun , Liu Guoming , Cao Liangzhi , Chen Qichang
Published: April 2011

The spherical harmonics (Pn) finite element method, the Sn finite element method, the triangle transmission probability method and the discrete triangle nodal method were all introduced to solve the neutron transport equation for unstructured fuel assembly respectively. The computing codes of each method were encoded and numerical results were discussed and compared. It was demonstrated that these four methods can solve neutron transport equations with unstructured-meshes very effectively and correctly, they can be used to solve unstructured fuel assembly problem.
Quantitative Evaluation of Hydrogen Micro-Damage in 316L Austenitic Stainless Steel
Authors: Toribio J. , Ayaso F. J.
Published: April 2011

In this paper an evaluation is made, by means of quantitative fractography and image analysis techniques, of the hydrogen-assisted micro-damage in axisymmetric notched samples of 316L austenitic stainless steel subjected to hydrogen embrittlement tests. Such a micro-damage created by hydrogen is concentrated in a external circumferential ring with the same centre as the cross sectional area of the notched samples. The microscopical appearance of this embrittled zone or damaged area is very rough and irregular at the micro-scale, with evidence of secondary cracking, in contrast with the smooth surface (at the micro-scale) created by micro-void coalescence (dimpled fracture) in the inner core which is not embrittled by hydrogen and fails by mechanical reasons. In addition, differences are observed in the matter of the appearance of the hydrogen-assisted micro-damage area as a function of the notched geometry and of the embrittlement time.
Design and Analysis of a Moderator Cell Structure for CARR CNS
Authors: Sun Shuo , Li Jian , Feng Quanke , Yu Xiaoling
Published: April 2011

A new moderator cell structure based on the single-phase thermo-siphon loop is proposed for the Cold Neutron Source (CNS) of China Advanced Research Reactor (CARR). Different from previous single-phase moderator cells, the design scheme of a crescent-shape hydrogen layer together with a helium cooling jacket is adopted in this paper. Theoretical analyses on cold neutron gain factor, heat transfer performance and structural strength of the moderator cell are carried out aiming at optimizing the preliminary design scheme. Results show that the modified moderator cell has high cold neutron gain factor, good heat transfer capacity and low stress, which could meet the function requirements of the CNS of CARR.
A Method of Optimized Utilization of Point-Wise Data Format in Monte Carlo Code
Authors: Liu Yuxuan , Yu Ganglin , Wang Kan
Published: April 2011

Monte Carlo codes are powerful and accurate tools for reactor core calculation. Most Monte Carlo codes use the point-wise data format, in which the data are given as tables of energy-cross section pairs. When calculating the cross sections at an incident energy value, it should be determined which grid interval the energy falls in. This procedure is repeated so frequently in Monte Carlo codes that its contribution in the overall calculation time can become quite significant. In this paper, the time distribution of Monte Carlo method is analyzed to illustrate the time consuming of cross section calculation. By investigation on searching and calculating cross section data in Monte Carlo code, a new search algorithm called hash table is elaborately designed to substitute the traditional binary search method in locating the energy grid interval. The results indicate that in the criticality calculation, hash table can save 5%∼17% CPU time, depending on the number of nuclides in the material, as well as complexity of geometry for particles tracking.
Numerical Research on Oscillation of Two-Phase Flow Under Rolling Motion
Authors: Zhang Y. J. , Su G. H. , Qiu S. Z. , Li H.
Published: April 2011

Two-phase flow instability and dynamics of a parallel multichannel system has been theoretically studied under periodic excitation induced by rolling motion in the present research. Based on the homogeneous flow model considering the rolling motion, the parallel multichannel model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The influences of the inlet, upward sections, and heating power on the flow instability under rolling motion have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained. The unstable regions occur in both low and high equilibrium quality and inlet subcooling regions. The multiplied period phenomenon occurs in the high equilibrium quality region and the chaos phenomenon appears on the right of MSB. The concept of stability space is presented.
Human Factors Engineering Verification and Validation: A Case Study of a Nuclear Power Plant
Authors: Song Fei , Zhang Shuhui
Published: April 2011

Main control room in nuclear power plants (NPP) is a complex system where operators interact with a large amount of human system interface (HSI) resources, which is essential to the safety of the plant. In order to achieve a high standard of human factors engineering (HFE) level and ensure the effectiveness and efficiency of the system, verification and validation (V&V) should be performed before the delivery of the plant. This study firstly represents an overview of the HFE project and V&V activities applied in the nuclear industry and relative researches, then, a V&V program from an ongoing 300MW NPP project is discussed in detail. Comparative methods, existing system vs. design requirements, are mainly applied for the verification phase. In the HFE Design Verification, 10 different HFE design guideline sets and corresponding checklists are developed for the experienced reviewers to conduct a thoroughness and consistency review. Critical safety functions, risk important tasks, critical human actions and other necessary operations are verified in the Task Support Verification to insure that HSI provides all required resources for personnel tasks. Other than direct comparison, a development platform is also used to assist the analysis of display, alarm and other appropriate features. In the validation phase, an integrated system test would be completed on a full scope simulator with the participation of experienced operators from the utility as subjects and multi-discipline observers. Scenarios are carefully designed to ensure the representative of the test. Both objective and subjective results would be collected and processed mainly by the descriptive statistics method to find out the problems in the existing design. All the discrepancies found in the whole V&V would be involved in a specific database and tracked until resolved. Lessons learned from this case are discussed.
Solids Formation Behavior of Simulated High-Level Liquid Waste During Long-Term Storage
Authors: Chen Jin , Liu Xuegang , Zhang Yanchao , He Qian’ge , Wang Jianchen
Published: April 2011

High-level liquid waste (HLLW) generated from reprocessing process contains actinides, lanthanides, fission products (FP) and a significant amount of nitrate ion. The partitioning and transmutation concept has been introduced for reducing the long-term hazards of HLLW. Several chemical separation processes mainly based on solvent extraction methods have been proposed to treat HLLW. However, solids consisting mainly Mo and Zr are known to form in HLLW during its long-term storage, Solid formations influence the composition of HLLW and the downstream solvent extraction process. To understand the precipitation behavior and stability of HLLW during its long-term storage, simulated HLLW (prepared as raffinate solution from LWR spent fuel reprocessing, 1AW solution) was prepared. Preliminary studies on solid formation behaviors with regard to the precipitation formation during refluxing and aging (representing a long-term storage) were carried out. Precipitation kinetics of major FPs such as Zr, Mo, Ru, rare earth elements, and etc. have been studied; The effect of phosphate ion concentration and temperature on solids formation were also experimentally examined. The formation conditions and the mechanism of solids were discussed.
The Mass Concrete Construction Methodology Exploring and Practice for Ensuring Quality Applied in the First AP1000 Project
Authors: Liu Xiao
Published: April 2011

With the research and development of nuclear power conception engineering and the requirement of structural integrity & durability, mass concrete structure was adopted more frequently in nuclear power construction project engineering. It caused that the construction requirements were more strictly on how to ensure construction quality of mass concrete for the nuclear power project which related to the nuclear safety at the first AP1000 plants in the world. Facing this challenge, exploring and practicing activities were taken on how to ensure mass concrete construction quality for the first AP1000 unit at Sanmen site with the considering of the quantity of concrete more than 2000 cubic meters with single pour seldom applied on nuclear power project. Based on the analysis and calculation factors which influence the mass concrete quality, the measurement and methodology system of mass concrete quality assurance was developed through the path of administration and technical. Success experience obtained and the preset quality realized through several mass concrete project construction activities on site. This thesis mainly introduce the detailed measurements explored and practiced for ensuring mass concrete construction quality at Sanmen nuclear power project construction site. The success experiences were generalized and summarized.
Performance Evaluation of Japanese Nuclear Power Plant Based on Open Data and Information (3)
Authors: Nagata Tadahisa , Sugiyama Ken-ichiro
Published: April 2011

The excessive maintenance of the nuclear power plants (NPPs) may cause the early (infant) failure in Japan. An easy analysis; the Weibull analysis was applied to the evaluation of the failure mode. The Weibull analysis needs the hazard data. The maintenance information of the equipment which caused plant shutdown was required for the hazard calculation. However, maintenance information of the equipment was not open. Therefore, all equipment was assumed to be maintained during every shutdown. This assumption was based on renewal process. However, a repair after unplanned shutdown of NPP is generally a restoration of only failed function without system overhaul. The system must be considered to age continuously. The system was not renewed. The operation data must be regarded as one continuous data before and after unplanned shutdown. An improvement of the Weibull analysis was required for NPPs. The model of the Weibull analysis was investigated. The competitive model in which shutdown caused by other than focused equipment/cause may be supposed to be continuous data could not be applied for a comprehensive analysis. Furthermore, the calculation method of the Weibull analysis was investigated. The calculation method of the hazard was viewed. A denominator of the hazard is the number of data which is cut for every continuous data by renewal process. However, multiple considerations of operation periods before unplanned shutdowns might cause underestimation of the failure rate in case of restoration process. Therefore, a dominator of the hazard was not supposed to be the number of data but the number of survived equipments (plants) at each time according to the definition of the hazard. This improved method is for the restoration process. The performance of Japanese NPPs was evaluated by improved method. The failure modes of Japanese NPPs were early failure modes. Moreover, performances of U.S. NPPs was tried to be evaluated by improved method. Operation data was collected from “NRC Power Reactor Status Reports”. However, many “maintenance outage”s which are the shutdowns of unknown origin were found. Therefore, DOE information was supplemented to investigate the “maintenance outage”. Failure modes of U.S. NPPs were the early failure modes, and failure rates were larger than Japanese NPPs.
Analysis of CHF Characteristics of Concentric-Tube Open Thermosyphon by Using Artificial Neural Network
Authors: Chen Ronghua , Su Gunaghui , Qiu Suizheng
Published: April 2011

An artificial neural network (ANN) for predicting critical heat flux (CHF) of concentric-tube open thermosyphon has been trained successfully based on the experimental data from the literature. The dimensionless input parameters of the ANN are density ratio, ρl/ρv, the ratio of the heated tube length to the inner diameter of the outer tube L/Di, the ratio of frictional area, di/(Di + do), and the ratio of equivalent heated diameter to characteristic bubble size, Dhe/[σ/g(ρl−ρv)]0.5, the output is Kutateladze number, Ku. The predicted values of ANN are found to be in reasonable agreement with the actual values from the experiments with a mean relative error (MRE) of 8.46%. For a particular outer tube, the CHF increases initially and then decreases with increasing inner tube diameter, and has a maximum at an optimum diameter of inner tube (do,opt). The do,opt is correlated with the working fluid and may decrease with the increase of ρl/ρv. CHF decreases with the increase of L/Di, and the decreasing rate decreases as L/Di increases. In the influence scope of pressure, the CHF decreases with increasing pressure for R22, while increases with increasing pressure for R113.
The Sensitivity Improvement of Passive Shutdown System Based on Curie Point Alloy in a Fast Reactor
Authors: Fan Nianqing , Lu Daogang , Zhu Minggang , Fang Yikun , Zhou Peide , Xie Chun
Published: April 2011

The magnetic properties will be lost when the temperature of Curie point alloy is above the Curie point temperature. The physical properties of Curie point alloy are used for Passive Actuated Shutdown System (PASS) in the Fast Reactor, for example American, Japan, the shutdown systems are called Self Actuated Shutdown System (SASS). The shutdown device is composed of the up assembly and the down assembly. The up one is a magnetic material and the down one is Curie point alloy. The reliability of PASS in normal condition, and whereabouts of sensitivity are determined by the properties of Curie point alloy. The Reactor-type Experiment Furnace and related circuits were designed to carry out static and dynamic experiments under the high temperature. Cylindrical alloy functions were investigated through the high temperature experiments outside of the reactor. The results showed that improvement was necessary because of long response time. In order to improve the sensitivity of PASS, different shapes were designed for Curie point alloy, under the premise that PASS was reliable. The response times of different sharps were calculated by finite element software, ANSYS, in abnormal condition. Comparison of response times of cylinder, quincunx and fins model under the same conditions, quincunx structure had the shortest response time when the temperature of Curie point ally was above the Curie point temperature. The results obtained from the high-temperature experiments, also showed that the quincunx has the shortest response time. These results showed the quincunx structure was superior to the other models.
APROS Analysis of Loviisa Sump Strainer Study
Authors: Kantee Heikki , Kontio Harri
Published: April 2011

The two Loviisa VVER-440 type reactors were commissioned in 1977 and 1980. The original designed life time of the reactors was 30 years. In 2003 Fortum, the owner and the operator of the Loviisa plant, launched an extensive safety study to prove the authorities that there was not any major safety issue why operating license could not be extended for another 20 years. In 2007 the Ministry of Employment and the Economy of Finland granted 20 and 23 years extension to the operating license for units 1 and 2, respectively. One issue, which needed further investigation, was the core cooling capability during sump circulation; i.e. were the present sump strainers good enough to prevent insulation fiber from not clogging the core coolant flow? Back in the 1990’s the original steel wire type sump strainers were replaced with stronger steel pipe type strainers. Some time later experiments were carried out to find out if insulation fiber could penetrate through the strainer holes and reduce the coolant mass flow rate through the core. The experiments indicated that the insulation fiber mixed with coolant partly penetrates through the strainer and gathers to the fuel assembly spacer grids increasing pressure loss across the core. The experiments were carried out in a rather simple test facility and also under forced single phase circulation. In those loss-of-coolant accidents (LOCA) where sump circulation takes place, circumstances are completely different. Therefore, it was decided that the APROS (Advanced PROcess Simulation) simulation software would be used to study the insulation fiber effect on core coolability during the accident. A large LOCA was chosen for the case to be analyzed. The reason for this was that during a large LOCA sump circulation begins in the early phase of the accident and a lot of emergency core cooling (ECC) water is injected into the primary circuit during sump circulation. The paper will first shortly discuss APROS simulation software. Then the test facility and the experimental results will be presented. The main issue is the analyses results. Several analyses were carried out to be able to determine the maximum amount fiber gathered in the spacer grids which the core can tolerate without overheating.
Preparation of Al2O3 Coated SiC Powders by Heterogeneous Precipitation
Authors: Zhang Dan , Suo Jinping
Published: April 2011

Alumina has been identified as a most promising material in tritium permeation barrier. By introducing some additive into the alumina improves its effectiveness. The purpose of this study was to synthesize Al2O3 coated SiC particles by heterogeneous precipitation method to replace the pure alumina as the raw material of coating. A series of factors which influenced the properties of coated powders had been studied, including the concentration of Al(NO3)3 solution, surface active reagent, the addition amount of SiC, calcinations temperature etc. The coated powders were analyzed by scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), X-ray diffraction (XRD) and Zeta-potential analyzer. The result indicates that the addition of certain amount of nano powers can reduce the phase transformation temperature of the alpha alumina. α-Al2O3 coated SiC powders were obtained by calcining Al(OH)3/SiC precursor at 1100 degree centigrade.
Status of “Low Power and Shutdown PRA Methodology Standard ANSI/ANS-58.22”
Authors: Wakefield Donald J.
Published: April 2011

The American Nuclear Society is developing a standard titled “Low Power and Shutdown PRA Methodology, ANSI/ANS-58.22.” It has been under development for more than 10 years now. During this time, drafts of the standard have been balloted by the ANS American Nuclear Society (ANS) Risk-Informed Safety Committee (RISC) on three different occasions; i.e., in 2005, 2006, and 2008. The most recent of these ballots occurred in November 2008; i.e., Reference 1. Each ballot failed to achieve a consensus for approval and numerous comments were received for improvement. Since the completion of the ballot on Version 8C in 2008, the Low Power and Shutdown (LPSD) writing group has been working on a volunteer basis to respond to the approximately 600 comments received and to revise the Standard for future ballot. A revised standard was completed in early November 2009. This latest draft was submitted in November 2009 to the ANS RISC pending a formal ballot for informal review. The intention of this informal review is to ascertain if members of the balloting RISC committee could support the revised standard and if not, to identify those issues for further work by the writing group. For approval of ANS standards, a consensus for approval must be received among all stakeholder groups, rather than a simple majority to approve.
The Prediction of Startup Test With Load Rejection and Parametric Study of SBPCS for Lungmen Nuclear Power Plant
Authors: Tsai Chiung-Wen , Shih Chunkuan , Lin Hao-Tzu , Wang Jong-Rong , Yuan Yng-Ruey , Cheng Su-Chin , Lin Fong-Lun
Published: April 2011

A Lungmen RETRAN-3D model has been constructed to predict the transient behaviors for startup test, furthermore verify the acceptance criteria specified in the documents of startup test procedure. This study focuses on the prediction of the startup test with Load Rejection (LR) with bypass and the parametric analysis of lead-lag time constants in pressure regulator. For the analysis of LR with bypass, the major mitigation functions, i.e., Selected Control Rods Run-In (SCRRI) and turbine bypass function, are simulated to examine whether scram is initiated during the transient or not. The analytic results show the reactor is brought to a steady state without scram. The neutron flux in the final state is around 34%, and the pressure regulator sensed maximum pressure rise is limited to a maximum of 3kPa. The result also shows that the 110% steam bypass capacity is capable to mitigate the power increase caused by the positive reactivity insertion as a result of pressure-wave-induced void collapse. For the parametric analysis of lead-lag time constants in pressure regulator, the time domain response of Steam Bypass and Pressure Control System (SBPCS) is demonstrated by a step change of pressure setpoint and different combinations of lead-lag time constants defined in pressure regulator. The results show that the responses, i.e., response time and overshooting, are minimized when the lag time constant is between 4 to 6 seconds and the lead time constant is 50% to 70% of lag time constant. The analysis result of SBPCS provides the trend as a reference for the adjustment of lead-lag time constants during the future Lungmen startup test.