bluuebunny/crossref_metadata_2025 huggingface dataset:
Stress Corrosion Cracking Incidents and Repair Technologies on PWR Dissimilar Weld Metal Joints in Japan
Authors: Chigusa Naoki , Hirano Shinro , Sera Takehiko , Kaguchi Hitoshi , Mukai Masayuki , Yoshida Yusuke
Published: April 2011

Several Japanese PWR power plants have experienced Primary Water Stress Corrosion Cracking (PWSCC) on dissimilar weld joints since 2004. J weld of 3 Reactor Vessel Head Penetration in Ohi unit 3 is one of the PWSCC incidents occurred in 2004 and has been studied by sampling and opening the fracture surface after its repair. Including Ohi unit 3 Reactor Vessel Head Penetration repair, Japanese PWR utilities and MHI have been developing the preventive maintenance and repair technologies applicable to alloy 600 welds and base metal, following PWSCC events on the Bugey-3 and V.C. Summer. This paper describes recent Japanese PWSCC incidents and repair technologies developed in Japan.
The Reliability Analysis of Bearing in Large Blower System
Authors: Zhao Jun , Zhang QinZhao , Tong JieJuan , Wang Hong
Published: April 2011

A large blower is one of the important subsystems in some complex gas transport. Typically, the blower system in the gas-cooled reactors act as the main pump in the water cooled reactors. The bearing is an essential equipment of the blower. In the vertical blower system, electromagnetic bearing and auxiliary bearing can be selected to suspend the blower system. Then the auxiliary bearing will bear the weight and kinetic energy of blower in the case of the electromagnetic bearing fails in process of blower running. Therefore, the reliability of the suit of bearing is one of the key questions in the design of blower. According to the design, the reliability of the suit of bearing is depended on the reliability of electromagnetic bearing and the design life of auxiliary bearing which is indicated by the times it can bear blower’s decline in the duration of running. In this paper, the reliability of this kind of bearing will be analyzed through a case study and the risk that it contributes to the blower system will be studied. The method of analysis is based on the reliability distribution of bearing and the Poisson distribution model. The advice about the design of bearing will be given from the results of analysis.
Improvement of the RIMS Detection Sensitivity for Application to Failed Fuel Detection and Location System of the Fast Reactor
Authors: Iwata Yoshihiro , Harano Hideki , Ito Chikara , Aoyama Takafumi
Published: April 2011

In the fast reactors, rapid and accurate detection of fuel failures as well as subsequent identification of failed fuel location are essential to achieve their safety operation and high plant availability. The gas tagging method, currently employed in the prototype fast breeder reactor Monju, is one of the efficient ways for the failed fuel detection and location (FFDL) technique, the principle of which is the isotope analysis of the argon (Ar) cover gas that includes, in case of fuel failure, a partial amount of leaked krypton (Kr) and xenon (Xe) originally loaded into each fuel pin. We propose a new type of FFDL technique using laser resonance ionization mass spectrometry (RIMS) for the isotope analysis of the cover gas in view of selective ionization of a specific element to obtain high S/N ratio. Nevertheless, the actual experimental data shows the existence of Ar and Ar2 non-resonant ionization by the photoelectron generated in the vacuum chamber to hinder precise measurement of Kr and Xe. We could successfully decrease the effect of these ions by one to two orders of magnitude by applying both a set of a neutralization apparatus and a Brewster window, and an electrode with a slit-type hole in the ion acceleration region, resulting in reliability improvement of RIMS in the FFDL system.
Progressive Evolution of the Core of the Fast Breeder Test Reactor
Authors: Varatharajan S. , Sureshkumar K. V. , Kasiviswanathan K. V. , Srinivasan G.
Published: April 2011

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.
Heat Transfer Characteristics of a Circular Channel Inserted Metallic Wire With High Porosity
Authors: Tendo Mitsutoshi , Takeda Tetsuaki
Published: April 2011

There are several methods for heat transfer enhancement. For example, there are attaching various fins on the heat transfer surface, processing the surface roughly, inserting twisted tape, and so on. These methods increase heat transfer coefficient or area by manufacturing of the heat transfer surface. However, it has to take into consideration the deterioration of the structure strength by attaching the fins on the tube surface with the design of the heat exchanger. The objective of this study is to clarify characteristics of heat transfer and pressure drop in the channel inserted metallic wire with high porosity. A heat transfer experiment has been performed using a horizontal circular tube to obtain the heat transfer characteristics in the channel inserted copper wire. This paper describes the heat transfer and flow characteristics of a heat exchanger tube filled with a high porous material. Fine copper wire (diameter: 0.5 mm) was inserted in a circular tube dominated by thermal conduction and forced convection. Working fluid was air. Hydraulic equivalent diameter was cited as the characteristic length in Nusselt number and Reynolds number. From the results obtained in this experiment, it was found that an amount of heat transfer in the tube with the copper wire was larger than that without one. An effectiveness of heat transfer enhancement increased with the temperature of the heated wall. The amount of heat transfer in the circular tube inserted copper wire, which has 0.993–0.998 of porosity, increased about 15% comparing with the tube having a smooth wall surface under the condition of the constant heat flux and lower than 170°C of the wall temperature.
Direct 3D MOC for Whole-Core Transport Calculation
Authors: Liu Zhouyu , Wu Hongchun , Cao Liangzhi , Chen Qichang , Li Yunzhao
Published: April 2011

Due to the need for high accuracy and more complex core geometry, whole-core transport calculations with the MOC are demanded. Recently, a 2D+1D method has been proposed to perform 3D whole-core transport calculations, but it is not easy to apply this method to the cases with complex geometry in the axial direction. Direct 3D MOC calculations are thus required. This approach is however limited by the amount of computer memory which is required for the ray tracing data that is required by the MOC. This paper presents a modular tracing technique and corresponding angular quadrature sets, which can greatly reduce the amount of ray tracing information, making it practical for the MOC to deal with the large 3D problems. In this paper, we introduce the model of the modular ray tracing technique and give some results of our method.
Neutron Transport Solution Using the Daubechies’ Wavelets in the Spatial Discretization
Authors: Zheng Youqi , Wu Hongchun , Cao Liangzhi
Published: April 2011

This paper describes a one-dimensional wavelet-based spatial discretization scheme for the first-order neutron transport equation. Two special features are introduced: i) the spatial variable is discretized using the Daubechies’ wavelets on the interval, and the neutron flux is represented in term of the wavelet series in a normalized node, the tradition SN angular discretization scheme is used in solving the equation, and ii) the wavelet Galerkin method is applied here, using the Daubechies’ scaling function as both the trialing function and weighting function, the integrations of Daubechies’ scaling function and its derivative in the Galerkin system are calculated numerically, using the difference quotient instead of the derivative. The boundary conditions and interface conditions are given in the exact form of wavelets series and added into the Galerkin system in special locations. The LU decomposition method is applied to solving the matrix in formed in the Galerkin system. The test results on several benchmark problems indicate that the wavelet-based spatial discretization scheme in this paper is capable of handling the first-order neutron transport equation, accurate in treating the boundary condition while using the wavelets expansion in spatial discretization, effective in treating the transport problems in the deep penetrating medium and in strong heterogeneous medium.
Quality Control of Interface Between Tungsten Coating and Cu Alloy Heat Sink of PFC for EAST Tokamak by Infrared Thermography
Authors: Qi Pan , Luo Guangnan , Li Qiang , Kuang Xiaojun , Zhou Haishan
Published: April 2011

Tungsten coated CuCrZr prepared by vacuum plasma spraying (VPS) has been chosen as Plasma-facing components (PFC) for EAST (Experimental Advanced Superconducting Tokamak) device, which have to survive from impact of high heat flux (HHF), fierce bombardment of particle flux and influence of alternating magnetic field in the device. Thus effective non-destructive examination (NDE) methods must be developed to assure quality and reliability of the PFCs. This paper presents infrared (IR) thermography NDE techniques being developed at ASIPP. Three-dimensional (3D) thermal finite element (FE) analyse has also been performed to verify the accuracy of the experimental results and improve the testing, showing that there is basic correlation between simulation and IR thermography testing.
Exponential Function Expansion Nodal Diffusion Method
Authors: Li Yunzhao , Wu Hongchun , Cao Liangzhi , Chen Qichang
Published: April 2011

An exponential function expansion nodal diffusion method is proposed to take care of diffusion calculation in unstructured geometry. Transverse integral technique is widely used in nodal method in regular geometry, such as rectangular and hexagonal, while improper in arbitrary triangular geometry because of the mathematical singularity. In this paper, nodal response matrix is derived by expanding detailed nodal flux distribution into a sum of exponential functions, and nodal balance equation can be obtained by strict integral in the polygonal node. Numerical results illustrate that the exponential function expansion nodal method in rectangular and triangular block can solve neutron diffusion equation in regular and irregular geometry.
ATHLET/BIPR-VVER Results of the OECD/NEA Benchmark for Coupled Codes on Kalinin-3 NPP Measured Data
Authors: Nikonov S. , Velkov K. , Pautz A.
Published: April 2011

The paper presents the results of the OECD/NEA benchmark transient ‘Switching off one main circulation pump at nominal power’ analyzed as a boundary condition problem by the coupled system code ATHLET-BIPR-VVER. Of primary interest are the comparisons done for the local in-core parameters — assembly outlet coolant temperatures at 93 measured points and SPND powers at 7 layers of 64 fuel assemblies. Revealed are some sources of inaccuracy and methods are proposed to be decreased. An important step is done for the future performing of uncertainty and sensitivity analysis in the frame of the OECD/NEA activities.
The Energy-Saving Diagnosis of PWR Nuclear Power Station Based on the Thermo-Economic Analysis Model
Authors: Chen Juan , Zhou Tao , Ran Ke
Published: April 2011

Exergy analysis model of PWR nuclear power station is developed in which signal flowing graph theory is introduced to set up the relation equations between input exergy flow and output exergy flow. Then, combining with resource distribution between different components, thermo-economic analysis model is obtained by setting up unit thermo-economic cost equations of different components with productive structure graph. Taking Daya Bay as an example, exergy analysis and thermal-economic analysis are put forward with detailed distribution of exergy and investment cost. Finally, aimed at energy-saving, static diagnosis is performed in two levels: energy conservation and cost reduction, and on this basis dynamic diagnosis is developed through sensitivity analysis considering different influence factors such as main steam temperature, fuel price, construction capital investment, post treatment cost and so on. The introduction of signal flow graph theory and thermal-economic structure theory is helpful to do performance estimation with high speed and good accuracy. It provides a new way for rapid optimization and offers an effective theoretical method for energy-saving of PWR nuclear power station including advanced reactor such as AP1000.
Estimation of SG TSP Blockage: Innovative Monitoring Through Dynamic Behavior Analysis
Authors: Midou Mickael , Ninet Julien , Girard Alexandre , Favennec Jean-Me´laine
Published: April 2011

EDF operates a fleet of 58 Pressurized Water Reactors (PWR). The “health” of the Steam Generators (SGs) is an essential element contributing to the overall thermal efficiency of a PWR, and finally to the availability of the unit. Among the health issues that may affect SGs, secondary-side corrosion products transport in PWRs may lead to many problems: various contaminants, both particulates and dissolved species, will unavoidably accumulate and concentrate in the Steam Generator. One consequence is the fouling of the heat transfer and support structure interfaces within the SG on the secondary side, especially the U-tubes (fouling deposits on the outer walls of the U-tubes), and the tube support plates (TSPs) that support the U-tubes. The accumulation of the corrosion products may lead to 3 main safety risks that must be monitored: fluid-elastic instability of tubes in flow-accelerated areas, a reduction in SG water mass inventory and an increase in the risk of water level oscillation. It has also significant performance issues because of the decision to power derate of some EDF PWRs. Thus, a global strategy to monitor the fouling and TSP blockage issues and to schedule preventive and curative actions has been designed and is under deployment by EDF nuclear operator. This dedicated periodic test relies on the recording of the following measurements in stabilized configuration: steam pressure, feedwater flowrate and temperature, primary circuit temperatures, SG blowdown flowrate and SG water level (wide and narrow range). A more precise monitoring of potential TSP blockage situations would be an interesting help to operation and maintenance strategies: deposit build-up in TSP foils could be minimized, preventive chemical cleaning operations could be scheduled and a more efficient fleet wide SG Management Program (SGMP) could be designed in accordance with secondary side deposit issues. Consequently, EDF R&D is experimenting a new method based on modeling dynamics behavior of SGs to assess a spatially distributed estimator of the TSP blockage ratio. This method, based on a 1D physical model of the SG that simulates the complex dynamics of the two-phase flow phenomena inside the SG, consists in computing the wide range water level responses according to various configurations during a particular transient which is particularly sensitive to this phenomenon. The TSP blockage ratio estimator is then obtained by comparing the computed response curves to those measured on-site. This new method has the potential advantages of being fully non-invasive, of providing a quarterly update of the TSPs blockage estimator, and of requiring no additional measurements by processing available plant data. It is also capable of estimating the efficiency of a chemical cleaning after restarting the plant and checking the evolution and kinetics of eventual TSP re-blockage.
Development of 1.2MV Induction Cavity for IVA Driver
Authors: Cong Peitian , Qiu Aici , Wu Hanyu , Zhang Guowei , Sun Jianfeng , Li Jingya , Su Zhaofeng
Published: April 2011

The 1.2MV, 70ns FWHM induction cell is developed for a 3MV Induction Voltage Adder (IVA) accelerator with three series connected cavities through a high voltage, vacuum insulating transmission line (VITL) driving rod-pinch diode (RPD) for radiography. The experimentally measured maximum relative permeability of IVA used amorphous material under pulse excitation is consistent with saturation wave model with flux density changing rate dB/dt greater than 10T/μs up to 32 T/μs. The remanence ratio Br/Bs is experimentally measured for IVA pre-annealed amorphous cores. Upon experimental results, cores in each cell are designed and determined according to current transfer efficiency and volt-second integral necessary for each induction cavity, and peak field preventing VITL bore negative surface from electron emission. The field analysis is carried out for designing VITL vacuum stack and oil cavity that contains cores and azimuthal transmission line with operation field stress being about 50% critical breakdown field. The prototype cell tests validate electric field safety and magnetic core performance. With magnetization inductance and resistance of eddy current calculated for inductive cell, the IVA accelerator circuit model is set up, and simulation predictions approximately according with experimental results are presented.
Introducing Nuclear Power Plants in an OECD Country: Size Influence on the External Factors
Authors: Locatelli Giorgio , Mancini Mauro , Galli Alessandro
Published: April 2011

A traditional investment evaluation approach is generally closed with a financial lifecycle performances investigation based on multiple analysis of Discounted Cash Flows (DCF). The international literature is rich of studies about the economics of new Nuclear Power Plants (NPPs), considering the classical accounts related to Construction, Operation & Maintenance, Fuel and Decommissioning. Financial analyses are important but the evaluation of such projects needs a multidimensional approach: besides economics, other technical, social and market factors have to be taken into account too. The Integrated model for the Competitiveness Assessment of SMRs (INCAS), developed by “Politecnico di Milano” cooperating with the IAEA, is designed to analyze the choice of the better NPP size as a multidimensional problem. The INCAS model aims to become the framework for the comparison between “Deliberately Small Medium Reactors (SMRs)” and “Large Reactors (LRs)”, with respect to a specific country situation or “scenario”. In particular the INCAS’s module “External Factors” evaluates the impact of factors not considered in the traditional DCF methods (siting and grid constraints, impact on the national industrial system, etc…) but critical for the decision of the plant size to deploy. This paper presents a completely updated version of the “External Factors model” framework under development since 2007. First it presents each factor, providing an adequate background and the quantification procedure, and then each factor is quantified respect to the Italian case. The IRIS reactor has been chosen as SMR representative. Even if the results are related to the Italian situation, they can apply to most of the European countries and the framework of the model can be used for all the countries. The results show that SMRs have better performances than LRs with respect to the external factors, in general and in the Italian scenario in particular.
Evaluation of the Conservatism of NPP Safety Analysis Dose Calculations as Typical for Licensing Purposes
Authors: Arnold Nikolaus , Mu¨llner Nikolaus , D’Auria Francesco , Mazzantini Oscar
Published: April 2011

Providing a reliable upper limit of radiological consequences to the plant personnel and the general public is typically the aim of a safety evaluation for anticipated operational occurrences or design basis accidents, as presented in a safety analysis report. A typical tool for dispersion calculation and dose evaluation is MACCS2. In the present analysis four types of calculations are presented: a first calculation, typical for licensing analysis, with the MACCS2 computer code. In a second step conservative assumptions e.g. ground release even if a stack release would be realistic, are dropped. In a third step calculation two is repeated with RODOS, a code (online decision making tool) used to predict the radiological consequences of an accidental release of activity. The step three calculation still contains all the conservative assumptions that are built in the MACCS2 code. In a last step these assumptions are removed, and a “best estimate” calculation on the dose to the public is performed. The whole analysis (step one to four) is repeated for different source terms (noble gases only, tritium dominated, primary system water …) and for different weather conditions. Two main conclusions can be drawn. The first by comparing step two (MACCS2) and step three (RODOS). Here the boundary conditions of the calculations are set to be as similar to each other as possible. The paper shows that despite the fact that MACCS2 uses a Gaussian plume model, while RODOS uses a puff model for dispersion calculation, doses of the same order of magnitude are calculated. For the second conclusion the step one (MACCS2, conservative) and step four (RODOS, best estimate) calculations are compared, it is shown that although the margin of conservatism varies considerably from case to case, the results differ at least one order of magnitude.
Sealing Solutions for an Aging Reactor
Authors: Sykes Timothy M.
Published: April 2011

There are an increasing number of nuclear reactors around the world operating well beyond their original design lives. Some pump seals supplied by original equipment manufacturers are either no longer available or their performance does not meet current requirements. This paper describes both the testing and operational experience of two nuclear pump seals that have been developed by AECL to replace the original equipment manufacturer’s seals in AECL’s National Research Universal reactor, which has been operating since 1957. The two seals described are the main heavy water pump seal and the heavy water degassing pump seal. The main heavy water pump seal is a tandem seal containing two identical, but 180° offset, eccentric seals of 95 mm balance diameter. The requirements were low pressure (0.3 MPa), low leakage (< 0.2 mL/min) and long, reliable lifetime (seven years). The seals were arranged to put full pressure across the outboard seal initially, with the inboard seal providing complete back up in the event of an outboard seal failure. Laboratory testing and operational performance, since initial installation in 1993, is described. The heavy water de-gassing pump seal design is based upon the main heavy water pump seal (a tandem seal containing two identical, but 180° offset, eccentric seals) but its balance diameter of 65 mm is much smaller than that of the main heavy water pump seal. Laboratory testing and operational performance, since initial installation in 2001, is also described.
Measurement of Thermal Neutron Energy Spectrum and Fluence of the Thermal Column in Xi’an Pulse Reactor
Authors: Zhong Yunhong , Jiang Xinbiao , Yu Qingyu , Zhang Wenshou
Published: April 2011

Neutron energy spectrum and fluence of the devices in a reactor are important parameters for the users. In this paper, the thermal neutron spectrum of the thermal column in Xi’an pulse reactor were measurement by the method of time of flight (TOF), and the peak of the energy spectrum is 0.0248±0.0006eV and the neutrons average energy is 0.048±0.001eV. With the spectrum results, the average cross section of 235U(n,f) and 197Au(n,γ) were calculated are respectively 92.08barn and 538.86barn. With the average cross sections, we measured the neutron fluence with the fission chamber and 197Au(n, γ) activation method. The measurements of the two methods are respectively 1.08×109n/m2s and 1.13×109n/m2s. We also calculated the standard uncertainty of the measurements: the energy spectrums’ is less than 5%, and the neutron fluence’ is less than 2.2%.
Self-Shielding and Criticality Calculation in Spent Fuel Dissolving Vessel
Authors: Liu Qingjie , Wu Hongchun , Cao Liangzhi
Published: April 2011

In the spent fuel dissolving vessel, the resonant isotopes in fuel pellets is surrounded by fissile solution. Complicated resonance interaction between the solid fuel and fissile solution is induced. Conventional self-shielding methods based on the equivalence theory for PWR cell and assembly will not apply because the neutron flux spectrum in the fissile solution are quite different from those PWR cases where no resonant isotope is in the moderator. In the present paper, Sub-group method is applied to calculate the space-dependent self-shielded multi-group cross-section. Sub-group parameters such as sub-group weights and sub-group cross-sections are calculated by fitting method using the NJOY processed multi-group constant under various dilutions. Elastic scattering Resonance for heavy isotopes is taken into account. Space-dependent sub-group flux is then calculated by fix source sub-group transport calculation. Spatially-dependent self-shielded multi-group cross-sections and multiplication factor can be obtained. Calculations are carried out on the NEA-CRP proposed standard problems for fissile pellets in fissile solution. Results show that the sub-group method is a very promising deterministic method for the self-shielding calculation in the spent fuel dissolution.
Design of Remote Control System for Far-Infrared Laser Interferometer
Authors: Shi Nan , Gao Xiang , Jie Yinxian , Wang Erhui , Yang Yao
Published: April 2011

A remote control system is designed to achieve the stable and economical operation of far-infrared laser interferometer on the EAST Machine (Experimental Advanced Superconducting Tokamak). Based on the PLC controller, the system is set up as two parts: gas flow control for adjusting working gas, feedback control for stabilizing output power of the laser. We demonstrate the feasibility that, according to the need of experiment, the working gas ratio mode can be changed in order to save gas (CD4) instead of keeping a single-mode ratio. Meanwhile, the changes of the output power and power supply of the laser have been measured due to the mode conversion. The output power of the DCN laser, as a laser source, is influenced by a number of parameters, especially the gas ratio and a variety of disturbances. Therefore, the feedback control system has been developed to make the laser power stable near the maximum available power. The whole system has been applied to the DCN laser, and has been proved to be effective.
Study on the Hydrogen Effect on the Temper Embrittlement of 2.25Cr-1Mo Steel
Authors: Zhang Xiliang , Zhou Changyu
Published: April 2011

The relationship between hydrogen embrittlement and temper embrittlement of 2.25Cr-1Mo steel was investigated by Auger electron spectrometer testing and electrochemical hydrogen charging. The results indicate that atomic hydrogen increases the grain boundary concentration of impurity element P and temper embrittlement. The increscent temper embrittlement degree by hydrogen will rise with further embrittlement and the relationship between increscent temper embrittlement degree and temper embrittlement degree is ΔCp = log[4.9351×(Cp)1.31687]. The grain boundary segregation theory was used and results of theory calculation and testing show that the relationship between hydrogen embrittlement and temper embrittlement can be explained by theory of grain boundary segregation. The embrittlement degree of 2.25Cr-1Mo steel which is induced by hydrogen and temper embrittlement is Cp′ = log[4.9351×(Cp)1.31687]+Cp.
A Study on Fault Diagnosis Technology of Nuclear Power Plant Based on Decision Tree
Authors: Mu Yu , Xia Hong
Published: April 2011

The technology of real-time fault diagnosis for NPP has great significance to improve the safety and economy of reactor. At present, expert system, artificial neural network (ANN) and support vector machine (SVM) algorithms are most widely used in the field of NPP fault diagnosis. According to the shortcomings of expert systems, ANN and SVM, the decision tree algorithm is applied in the field of NPP fault diagnosis in this paper. ID3 and C4.5 are applied separately to learn from training samples which are the typical faults of NPP, and diagnose using the acquired knowledge. Then the diagnostic results are compared with the results of SVM method. The results show that: comparing with SVM, decision tree has the advantages of much faster training speed and a little higher accuracy. Furthermore, decision tree can obtain rules from the sample set, so it has good explanatory ability for the diagnostic results.
The Design and Simulation of Voltage-Preamplifier With Numerical-Controlled Plus
Authors: Ma Wenyan , Yu Guojian , Chang Jian
Published: April 2011

Usually the plus of the voltage-preamplifier we used is changeless or 2-modes (×1 or ×10), and the value is low (×10). But in the practical application, the energy of the particles detected are complicated. Especially, when the particles which are detected are unknown, it will be advantageous to the result to select a proper plus. So it will be great beneficial for the practical application to design a bran-new voltage-preamplifier which has bigger plus comparing to the amplifier we used now and could select a proper plus for the experiment automatically. This article introduces a design of a kind of new voltage-preamplifier with a bigger plus (×20) than traditional voltage-preamplifiers and whose plus can be controlled continuously. The most important of all is that this voltage-preamplifier can select a appropriate plus for the experiment automatically by the order of the handlers. It adjusts the plus by the information from the sensor to the order. In this design we control the output-signal from 100∼200 mV. And then, this paper introduces the simulation of the amplifier’s working process by Multisim10.0 and Multisim10.0 Proteus ISIS.
Analytic Basis Function Expansion Nodal Method for Neutron Diffusion Equations in Triangular Geometry
Authors: Wang Kunpeng , Wu Hongchun , Cao Liangzhi , Wang Changhui
Published: April 2011

An analytic basis function expansion nodal method for directly solving the two-group neutron diffusion equation in the triangular geometry is proposed in the present paper. In this method, the distribution of neutron flux is expanded by a set of analytic basis functions. The diffusion equation is satisfied at any point in a triangular node for each group assuming that the flux within a node is flat. No transverse integration is needed. To improve the nodal coupling relations and computation accuracy, nodes are coupled with each other fulfilling both the zero- and first-order partial neutron current moments across all the three interface of the triangle mesh at the same time. Coordinate conversion is used to transform arbitrary triangle into regular triangle in order to simplify the derivation. A new sweeping scheme is developed for the triangular mesh and the response matrix technique was used to solve the nodal diffusion equation iteratively. Based on the proposed model, the code ABFEM-T is developed. Validation of code for accuracy and efficiency are carried out by calculating both rectangular and hexagonal assembly benchmark problems. Numerical results for the series of benchmark problems show that both the multiplication factor and nodal power distribution are predicted accurately. Therefore this method can be used for solving neutron diffusion problems in complex unstructured geometry.
Passive Safety Systems of Advanced Nuclear Power Plant: AP1000
Authors: Yan Guohua , Ye Chen
Published: April 2011

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.
Self-Healing Behavior of Pores in Atmosphere Plasma Sprayed Tritium Permeation Barrier Coatings
Authors: Gao Jifeng , Suo Jinping , Zhang Dan
Published: April 2011

Decrease of pores in tritium permeation barriers is one of the most important problems to be addressed for the proper functioning of the fusion reactor. In this paper, a self-healing composite coating composed of TiC+mixture (TiC/Al2O3) +Al2O3 was developed to solve this problem. The coating was deposited on martensitic steels by plasma spraying with a thickness of 100μm. After heat-treatment, the morphology and phase of the coating were investigated by scanning electron microscopy (SEM) and X-ray diffraction (XRD). The corrosion resistance of the coating before and after heat treatment was examined by electrochemistry techniques. The results showed that the TiC+mixture (TiC/Al2O3)+Al2O3 coating exhibited good adhesion to the substrate and a perfect self-healing ability with the porosity decreased by 90% after heat-treatment. The corrosion resistance of the coating increased evidently after the heat treatment. The oxidation/expansion of TiC in the coating played an important role in the sealing of pores.
Treatment of Neutron Cross-Section With Interpolation
Authors: Zhang Xiang , Yu Ganglin , Bi Guangwen , Wang Kang
Published: April 2011

Using NJOY to generate the temperature dependent neutron cross-section is too time-consuming in practice, especially for many nuclides. So an approach involving interpolation between nuclear data libraries at different temperatures is investigated. Based on the ACE data at different temperatures, we used ITND — an neutron cross-section interpolation program, to generate the target temperature ACE data, then we compared it with the ACE data which generated by NJOY at the same temperature. We focused on the interpolation result of 238U, 235U, 232Th, Zr, 16O, 10B and 1H at the temperature of 575K. To that nuclides, several interpolate schemes were studied, and we demonstrated the relative differences, and explain their reasons. Finally we applied these ACE data to benchmark calculation, and good agreement was observed with the benchmark results.
Characteristics of Cavitation and Erosion Phenomena in Sodium Flow
Authors: Ardiansyah Teddy , Asaba Makoto , Miura Kuniaki , Takahashi Minoru
Published: April 2011

This paper presents experimental data of cavitation experiments to determine the characteristics of cavitation and erosion in the flowing liquid sodium at 200–400°C. The test section is a venturi made from SUS316 since this material is used as the cladding material for SFR. The ID and OD of the venturi test section are 6.5 mm and 21.4 mm, respectively. The data show that the onset cavitation conditions (onset velocities) in liquid sodium are influenced by the change of the stagnant pressure at the expansion tank. Cavitation noise signals at developed cavitation conditions are not affected by the change of stagnant pressure and might be caused by choking of the sodium flow that restricts the formation of cavitation bubbles. For all the different stagnant pressures and temperatures, the onset cavitation coefficient K is around unity. Meanwhile, erosion experiment in the flowing liquid sodium for 600 hours at 200°C and cavitation coefficient value K of 0.59–0.51 (developed cavitation condition) reveals that cavity bubbles produce some micro pits at the outlet of the venturi test section. These pits indicate that erosion is occurred on the surface of the tested material and could be a major problem for the development of SFR if cavitation occurs inside the critical/important parts.
Structural Design of S-Bend Box for ITER Feeder Based on Optimization Technique of ANSYS
Authors: Dong Biao , Wang Kai-song , Song Yun-tao
Published: April 2011

On the premise that ensuring the reliability of structural strength, and in order to get the optimal structural parameters and the number of reinforcement rib of the ITER feeder S-bend box (SBB), so that its stress distribution is more uniform and reasonable under the critical pressure load, and the cost of materials is relatively smaller, this article, through theoretical calculations and inferences on the basis of the parametric modeling and static analysis and checking of SBB, setting up different reinforcement rib numbers, optimizes through the ANSYS optimization design module, with the maximum stress as the objection function, with and reasonable quality and displacement as state variables, with the size of wall thickness and the space of reinforcement rib as design variables. And it regulates true stress at stress concentration through submodeling of ANSYS and local solid modeling. The results of optimization analysis show that: SBB reached the optimal solution when the reinforcement rib number N of SBB takes 3, the maximum equivalent stress is 117 Mpa and the weight is 6417Kg. Finally, SBB structural parameters, which obtained through optimization design, are rounded according to GB. These meet the design requirements, correspond to the practical applications and provide technical parameters and basis for the future development of SBB.
The Consistency of WIMS-D Library and ACE Format Library
Authors: Liu Zhiyan , Wu Hongchun , Cao Liangzhi , Liu Qingjie
Published: April 2011

Nuclear data library is the cornerstone in the nuclear reactor’s design and calculation. The WIMS-D multi-group library and ACE format library (mainly used in MCNP) is applied frequently in the nuclear calculation. We have developed a new self-shielding calculation procedure based on Wavelets scaling function expansion method. This procedure needs several parts in both WIMS-D and ACE format library. So the consistency of two libraries becomes a very serious problem. This may bring in large errors. In this paper, NJOY cross section processing system is used to produce new WIMS-D and ACE format library from the same ENDF/B data. We compute some homogenous problems using new and old libraries in WIMS-D and ACE format. The results of the two new libraries and the old libraries are compared respectively. It is found that there are consistency problems between the two libraries. The newly produced libraries are more compatible than the old ones.
Localization of a Noise Source in VVER-1000 Reactor Core Using Neutron Noise Analysis Methods
Authors: Malmir Hessam , Vosoughi Naser , Zahedinejad Ehsan
Published: April 2011

In this paper, localization of a noise source from limited neutron detectors sparsely distributed throughout the core of a typical VVER-1000 reactor is investigated. For this purpose, developing a 2-D neutron noise simulator for hexagonal geometries based on the 2-group diffusion approximation, the reactor dynamic transfer function is calculated. The box-scheme finite difference method is first developed for hexagonal geometries, to be used for spatial discretisation of both 2-D 2-group static and noise diffusion equations. The dynamic state is assumed in the frequency domain which leads to discarding of the time disrcetisation. The developed 2-D 2-group neutron noise simulator calculates both the discretised forward and the adjoint reactor transfer function between a point-like source and its induced neutron noise, by assuming the noise source as an absorber of variable strength type. Benchmarking of the mentioned neutron noise simulator revealed that it works satisfactorily. Finally, by using the inversion method of reconstruction, the location and values of a noise source of the type absorber of variable strength (or reactor oscillator) in VVER-1000 reactor cores are determined. Accuracy of this method is highly acceptable.
Multi-Group Library Generation for the Study of Nuclide Transmutation in High Flux Engineering Test Reactor
Authors: He Lei , Wu Hongchun , Cao Liangzhi , Liu Qingjie , Liu Zhiyan
Published: April 2011

Transmutation of the nuclear waste which contains MA and FP is important and necessary. Because the major MA and FP’s nucleus densities is large in transmutation calculation, now existing multi-group WIMS libraries are not suitable for transmutation calculation. This paper makes a study for the parameters value choice such as weight spectrum and background cross-section σ0 in NJOY. With the new parameters value, produce suitable multi-group data of FP and a new library for transmutation calculation. In purpose of comparing these two libraries, build a simple transmutation cell model in HFETR. Achieve the cell calculation separately by WIMSD with new library and original library. Then compare the results of parameters such as K-INF, flux, and absorption reactions. The results show that the values of these parameters are different in resonance energy range, so it is necessary to think over the resonance of FP when a multi-group library is produced for thermal reactor transmutation.
Study on the Interfacial Behavior and Thermal Characteristics in Liquid-Vapor Flow of the Supersonic Steam Injector
Authors: Shibayama Shunsuke , Abe Yutaka , Kaneko Akiko , Fukuichi Akira , Iwaki Chikako , Narabayashi Tadashi , Mori Michitsugu , Ohmori Shuichi
Published: April 2011

Supersonic steam injector is a passive jet pump which operates without rotating power source or machinery and it has high heat-transfer performance due to the direct contact condensation between supersonic steam flow and subcooled water jet at the mixing nozzle. Since the supersonic steam injector has a quite simple and compact structure, it has been considered to apply to the safety system for the Next-generation nuclear power plant. There are various researches about the formulation and modeling of operating, flow structure and heat transfer characteristics of both vapor and liquid flow. However, there are few models which are capable of evaluating heat and momentum exchange at the boundary layer between supersonic steam flow and water jet. Since heat and momentum exchange is considered to have a major impact to operating characteristics of a supersonic steam injector, it is necessary to formulate the model which simulates such complex phenomena at the boundary layer with high accuracy. The objective of the present study is to investigate the relation between the thermal characteristics and interfacial behavior between the flows to develop a model which is able to assess the heat and momentum transfer characteristic and the flow structure of the supersonic steam injector in detail. In the present study, a visible test section of water jet-centered supersonic steam injector was adopted to conduct visualization of the water jet with high speed video camera. In addition, special measurement instrumentations of temperature and pressure were applied to obtain radial distribution of temperature and pressure in the mixing nozzle of the injector. There were large velocity and temperature gaps between the water jet and the supersonic steam flow which indicated the existence of large momentum and heat exchange at the boundary layer of the flows. It was clarified that there was pressure gradient which was considered to stabilize the water jet in the mixing nozzle from calculation of radial distribution of total pressure gradient. From the visualization measurement, it was also clarified the existence of a complex wavy behavior on the surface of the water jet. The wave velocity was estimated by the image processing technique and the cross-correlation method. It was found that there was a relation between the wave velocity and heat transfer characteristics in the supersonic steam injector. It is suggested that the enthalpy ratio between the liquid subcool enthalpy and steam condensation enthalpy as well as the Jacob number between both flows could be an indication factor for the effect of the wavy behavior on the condensation.