bluuebunny/crossref_metadata_2025 huggingface dataset:
Stury SBO Accident of Typical M310 NPP With Probabilistic Risk Assessment Method
Authors: Deng Wei , Feng Sun
Published: April 2011

The availability of alternating current (ac) electrical power is essential for the safe operation and accident recovery of commercial nuclear power plants (NPPs). And reliable off-site power is one key to minimizing the probability of severe accidents. This paper mainly issues the risk of SBO resulting from Loss Of Offsite Power (LOOP) on the typical M310 type stations (3-Loops, 1000MWe, pressurized NPP) of China during full power operation and normal shutdown/RHR states with constructing SBO event trees and fault trees, and Human Reliability Analysis (HRA) and data analysis are also taken into account in this analysis. Further more, Based on these work, the Core Damage Frequency (CDF) values and associated Minimal Cut Sets (MCSs) are gained. The results from the analysis give us some useful information that can help to find the weaknesses of NPP design or operation and to improve the safety level of NPPs operation.
General Overview of CVCS Design for the TSN Project
Authors: Ciriello Antonio , Liu Man
Published: April 2011

This paper resumes the results of the collaboration between AREVA and CNPDC during the past two years for performing and achieving the basic design of EPR™ reactor CVCS system for the TSN NPP. The CVCS (Chemical and Volume Control System) is an essential auxiliary system of the PWR technology based nuclear power plants all around the world. In the EPR™ reactor design, as it is also in similar nuclear power plants, this auxiliary system has well determined functions, which are: reactivity control, reactor coolant volume control, coolant chemistry control, primary system main pumps seal water injection as well as the pressurizer auxiliary spray regulation for the Reactor Coolant System. In the EPR™ reactor design, the CVCS is mainly an operational system and only some valves and instrumentations take part at some specific safety functions, (e.g. Reactivity Control, Containment of Radioactive Substances). In the first part of this paper a general introduction to the EPR™ reactor CVCS technology, including the related safety functions and detailed operational functions of CVCS, is presented. In the TSN EPR™ reactor CVCS design, the system is divided into eight sections, (defined from RCV1 to RCV8). The corresponding detailed description of these sections, including their functions, structure and main components, as they have been implemented in the EPR™ reactor CVCS design for the TSN NPP, is then presented in the second part of this paper. In addition some specific design features for EPR™ reactor CVCS system for the TSN NPP, such as the hydrogenation station technology, are also focused in this paper. The reference power plant, concerning the CVCS design, for the TSN NPP is the FA3 NPP, but different design concepts have been implemented in the TSN NPP with regards to the coolant purification section (RCV2), and the coolant filtering in the reactor coolant pumps seal injection and leak-off lines.
A Parametric Study of Fuel Lattice Design for HTR-10
Authors: Wang Meng-Jen , Peir Jinn-Jer , Chi Chen-Wei , Liang Jenq-Horng
Published: April 2011

In this study, the multiplication factor and neutron spectrum behaviors were investigated against the moderator-to-fuel ratio, the fuel loading height, and the detector location in HTR-10. The MCNP5 computer code (version 1.51) was employed to perform all the simulation computations. The results revealed that the multiplication factor varies significantly depending on the moderator-to-fuel ratio and the fuel loading height due to the competition among the neutron moderation and absorption abilities of the moderator as well as the neutron production ability of the fuel. Due to its inherent stability, HTR-10 is deliberately designed such that the multiplication factor decreases and the neutron spectrum softens as the moderator-to-fuel ratio increases. The average neutron energy level in the HTR-10 fuel balls is approximately 200 keV and ranges from smallest to largest at the middle, bottom, and top of the reactor core, respectively.
Validation of a New Core Simulator COSMO-S for PWRs
Authors: Tsubota Shinobu , Sato Daisuke , Yamaji Kazuya , Koike Hiroki , Matsumoto Hideki
Published: April 2011

A new PWR nuclear design code system, GALAXY/COSMO-S, has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) with support of Osaka University and Nagoya University. The code system has been developed based on today’s popular techniques such as MOC and semi-analytical nodal expansion method for GALAXY and COSMO-S, respectively. These codes employ several new features to improve calculation accuracy and efficiency of commercial PWR design calculations. Especially, a new robust cross section representation model based on the concept of quality engineering is introduced in GALAXY/COSMO-S system. The new representation model is intended to cover all PWR operating conditions ranging from cold shut down to hot full power. The robustness of the cross section representation model is achieved by fitting equation including cross term effect by fuel temperature, moderator density and moderator temperature. The fitting equation is determined based on Akaike’s Information Criteria (AIC) instead of empirical ways. It is the first challenge to apply the concept of quality engineering technique, which is widely used in manufacturing field, to the cross-section representation model. In order to demonstrate the applicability of COSMO-S for PWR core designs, several benchmark calculations simulating an assembly models, multi assembly models and a typical commercial plant were performed. Reactivity and power distribution were compared with the reference results prepared by GALAXY and current core design code. The results of these comparisons show excellent agreements with the references. These results validate the applicability of the GALAXY/COSMO-S system with a new cross section model for PWR core design.
A Model for Passive System Reliability Analysis
Authors: Yu Yu , Tong Jiejuan , Liu Tao , Zhao Jun , Zhang Aling
Published: April 2011

It is the important feature of passive system and the basic difference from the active system that nuclear plant can be driven to safe state or shutdown by inherent safety characters of the reactor and physical principles, independent of human interfere or the operation of outside equipments, when the reactor is in abnormal condition. So passive system is widely used in new generation nuclear power plant (NPP) such as high-temperature gas-cooled reactors and AP1000 NPPs. While physical process failure become one of the important contributors to the system operation failure since system operation is depending on natural force but not on outside power and both the driven force and resistance are influenced by many uncertain factors. Then finding the key factors for the system operation, analyzing the development of the passive system combining with the accident scenarios are the main steps of the analysis of the passive system reliability, and the important content of the probability safety assessment (PSA) of nuclear plant with the passive design. In this paper, a model for analyzing the passive system reliability is described, in which variance decomposition and analytic hierarchy process (AHP) methods are used to select the key factors for the system operation, and Monte Carlo simulation and dynamic event tree methods are used to evaluate the system reliability according to the accident scenarios. Finally, Passive Residual Heat Removal System in the High Temperature Gas-Cooled Reactor (HTGR) is analyzed as an example.
Study on Characteristics of Gas-Liquid Two Phase Flow Distribution in Vertical Manifolds
Authors: Sheng Cheng , Zhou Tao , Pang Liping , Huang Yanping
Published: April 2011

This paper shows the experiments of the flow rate distribution measurements of air-water two-phase flow in vertical multi-branch carried out on the Gas-liquid two-phase flow fluid and flow rate distribution experimental platform, and discusses the laws and the Characteristics of two phase flow distribution in the multi-branch by changing the water flow rate or the air flow rate under the condition of fix the flow rate of the other phase, then analyzes the degree of the flow deviation of both two phases in branches by calculating the standard deviations of each working condition. The experiments show that the flow distributions of both two phases are very uneven. There was much air but little water flows into the branches which were close to the inlets of the distributor, but in the branches which were far away from the inlets of the distributor, the situation was opposite to the previous one. The types of the flow pattern in each branch in every working condition were obtained through the Hewitt-Robert flow pattern picture. As the flow rate of air increased, the annular pattern would spread from the branches which are close to the inlets of the distributor to the ones which are far away from the inlets of the distributor.
Experimental Study on Fire-Extinguishing of Lithium
Authors: Furukawa Tomohiro , Kato Shoichi , Hirakawa Yasushi , Kondo Hiroo , Nakamura Hiroo
Published: April 2011

Fire-extinguishing behavior of four fire extinguishants, dry sand, pearlite, Natrex-L and Natrex-M on burning lithium was examined. Temperature and flame increase in chemical reaction between lithium and silicon, which is the major element in the fire extinguishants, were observed for dry sand and pearlite. For Natrex-L, temperature increase was not observed visually, although flame was slightly increased when it was applied to the burning lithium. The effect of lithium pool depth on the fire-extinguishing performance of Natrex-L was investigated on the definite area of the lithium combustion surface because the density of Natrex-L was larger than that of liquid lithium. It was found that the amount (thickness) of fire extinguishant necessary for fire-extinguishing increased as the depth increased. In this experimental condition (combustion area: 270cm2, lithium depth: 1–2cm), the minimum thickness of the fire extinguishant was 1.5 times the depth of the lithium pool.
Effective Software Verification and Validation Approach for Nuclear Power Plant Digital Instrumentation and Control Systems
Authors: Yang Steve , Ding Jun , Miao Huifang , Zheng Jianxiang
Published: April 2011

All 1000 MW nuclear power plants currently in construction or projected to-be-built in China will use the digital instrumentation and control (I&C) systems. Safety and reliability are the ultimate concern for the digital I&C systems. To obtain high confidence in the safety of digital I&C systems, rigorous software verification and validation (V&V) life-cycle methodologies are necessary. The V&V life-cycle process ensures that the requirements of the system and software are correct, complete, and traceable; that the requirements at the end of each life-cycle phase fulfill the requirements imposed by the previous phase; and the final product meets the user-specified requirements. The V&V process is best illustrated via the so-called V-model. This paper describes the V-model in detail by some examples. Through the examples demonstration, it is shown that the process detailed in the V-model is consistent with the IEEE Std 1012-1998, which is endorsed by the US Regulatory Guide 1.168-2004. The examples show that the V-model process detailed in this paper provides an effective V&V approach for digital I&C systems used in nuclear power plants. Additionally, in order to obtain a qualitative mathematical description of the V-model, we study its topological structure in graph theory. This study confirms the rationality of the V-model. Finally, the V&V approach affording protection against common-cause failure from design deficiencies, and manufacturing errors is explored. We conclude that rigorous V&V activities using the V-model are creditable in reducing the risk of common-cause failures.
Fracture Strength Estimation of SiC Block for IS Process
Authors: Takegami Hiroaki , Terada Atsuhiko , Onuki Kaoru , Hino Ryutaro
Published: April 2011

The Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production to meet massive demand in the future hydrogen economy. A concept of sulfuric acid decomposer was developed featuring a heat exchanger block made of SiC. Recent activity has focused on the reliability assessment of SiC block. Although knowing the strength of SiC block is important for the reliability assessment, it is difficult to evaluate a large-scale ceramics structure without destructive test. In this study, a novel approach for strength estimation of SiC structure was proposed. Since accurate strength estimation of individual ceramics structure is difficult, a prediction method of minimum strength in the structure of the same design was proposed based on effective volume theory and optimized Weibull modulus. Optimum value of the Weibull modulus was determined for estimating the lowest strength. The strength estimation line was developed by using the determined modulus. The validity of the line was verified by destructive test of SiC block model, which is small-scale model of the SiC block. The fracture strength of small-scale model satisfied the predicted strength.
Current Status of Design and Construction of IFMIF/EVEDA Lithium Test Loop
Authors: Kondo H. , Furukawa T. , Hirakawa Y. , Matsushita I. , Ida M. , Horiike H. , Kanemura T. , Sugiura H. , Yagi J. , Suzuki A. , Terai T. , Fukada S. , Nakamura H.
Published: April 2011

The International Fusion Materials Irradiation Facility (IFMIF) is a D+-Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF started on July 2007. Regarding to the lithium (Li) target facility, design and construction of EVEDA Li Test Loop is a major activity and is in progress. This paper presents the current status of the design and construction of EVEDA Li Test Loop. The EVEDA Li Test Loop consists of a main loop system and a purification loop system. The detail design was started at the early 2009. Fabrication of the loop was started at middle of 2009, and completion is planned at the end of Feb. 2011. Currently, the system diagram of the EVEDA Li Test Loop is finished to be defined. The diagram and function of major components in the main loop system and the purification loop system are described in this paper.
Recent Advances in Integrated Nuclear Heating Reactor Technology
Authors: Guo Jilin , Li Weihua , Zhang Yajun , Yu Ruixia , Ma Jizhong , Meng Dongwang
Published: April 2011

The recent advances in utilization of nuclear heating reactor technology are to couple with seawater desalination process, in which process steam and fresh water is combined to produce, steam turbine is used, thermal and membrane processes are joined. In this technical concept, the “zero discharge” to environment will be realized due to using nuclear energy as the heat source for process steam and seawater desalination. Combined process steam and fresh water produce will make the product’s cost lower and make the whole system economic results better. The special design features of nuclear heating reactor will ensure that any radiological contamination can be excluded from fresh water. As joining thermal process and membrane process seawater desalination, the step utilization of thermal energy and reuse of waste heat can be realized.
Development of System Based Code: Part 1—Reliability Target Derivation of Structures and Components
Authors: Kurisaka Kenichi , Nakai Ryodai , Asayama Tai , Takaya Shigeru
Published: April 2011

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code (SBC) by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission (NSC) of Japan and the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF) that were determined in the Fast Reactor Cycle Technology Development (FaCT) project by Japan Atomic Energy Agency (JAEA), by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor (JSFR). The risk from the reactor is expressed with sum of combination of various elements in the PSA analysis model. Those elements include not only static failure of the structures and components. However, the present study focuses on the sequences including the static failure, and the probability of dynamic failures and human errors in those sequences is conservatively assumed as a unity. It was confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.
Study of the Space-Time Neutron Multiplication Formula
Authors: Hao Jianli , Chen Wenzhen , Wang Shaoming , Zhang De
Published: April 2011

The process of neutron multiplication is a discrete-time process, but the neutron transport theory takes neutron multiplication as a continuous neutron source, which ignores the discrete-time process of neutron multiplication, which would take in errors, so it is necessary for describing the process of neutron multiplication as a discrete-time process. “The neutron doubling formula including delayed neutrons” has been established which describes the process of neutron multiplication as a discrete-time process, but it has nothing to do with space. “The neutron doubling formula including delayed neutrons” could not be used to describe the variety of distributing of neutron density in transient process; it also could not be used to deal with the problem of three-dimensional space. In order to solve the problems mentioned above, the space-time neutron multiplication formula is established. Based on the theory of neutron multiplication, the concept of space is introduced to the neutron multiplication formula and the space-time neutron multiplication formula is established by taking into account of neutron transport. The formula can describe the inherent physical process of neutron multiplication in fission chain reaction system. The test of space-time neutron multiplication formula is done, which proves the formula is right. Given the initial neutron density as well as the multiplication factor, the formula can strictly describe the variety of neutron density (neutron flux density) with time. It could be used for setting a standard for estimating error for the measurement of neutron flux density as well as numerical calculation; the space-time neutron multiplication has larger applicability compared with the “neutron doubling formula including delayed neutrons”.
Analysis of Single Phase Flow and Temperature Fields in Rectangular Channel With Translation in Width
Authors: Luo Lei , Chen Wenzhen , Chen Zhiyun , Hao Jianli
Published: April 2011

Plate fuel assemblies will be widely applied in the future because of their simple and compact structure and excellent heat exchange capability. When the plate fuel assemblies are used in a ship reactor, the effect of the ship motion on the reactor core thermal hydrodynamic must be investigated. In this paper, aiming at the rectangular shape of coolant channels in the assemblies, the numerical simulation of the flow and heat exchange in a rectangular channel in horizontal translation is carried out. Using the CFD Software and UDF code, the simulation model of the channel in horizontal translation with and without acceleration along its width direction is built up, and the related characteristics is analyzed and discussed. The results and conclusions are applicable and useful for the new type ship PWR design.
Test Facility Design for Integrated Digital Nuclear Reactor Protection System
Authors: Xiong Huasheng , Li Duo , Zhang Liangju
Published: April 2011

Reactor protection system is one of the most important safety systems in nuclear power plant and shall be designed with very high reliability. Digital computer-based Reactor Protection System (RPS) takes great advantages over its conventional counterpart based on analog technique and faces the issues how to effectively demonstrate and confirm the completeness and correctness of the software that performs reactor safety functions in the same time. It is commonly accepted that the essential way to solve safety software issues in a digital RPS is to pass a strict and independent Verification and Validation (V&V) process, in which integrated RPS testing play an important role to form a part of the overall system validation. Integrated RPS testing must be carried out rigorously before the system is delivered to nuclear power plant. The integrated testing are often combined with the factory acceptance test (FAT) to form a single testing activity, during which the RPS is excited by emulated static and dynamic input signals. The integration testing should simulate normal operation, anticipated operational occurrences and accident conditions, as well as anticipated faults on the inputs to the DRPS such as sensors out of range or ambiguous input readings. All safety function requirements of digital RPS should be confirmed by representative testing. The design and development of a test facility to carry out the integrated RPS testing are covered in this paper, which is merged in the research on a digital RPS engineering prototype for a nuclear power plant. The test facility is based on PXI platform and LabVIEW software development environment and its architecture design also takes into account the test functions future extensions such as hardware upgrades and software modules enhancement. The test facility provides the digital RPS with redundant, synchronized and multi-channel emulated signals that are produced to emulate all protection signals from 1E class sensors and transmitters with time varied value within their possible ranges, which would put integrated RPS testing into practice to confirm the digital RPS has fully met its predefined safety functionality requirements. The designed test facility can provide an independent verification and validation process for the research of digital RPS with scientific methods and authentic data to evaluate the RPS performance thoroughly and effectively, such as measuring threshold precision and trip response time, analyzing system statistical reliability and so on.
The Application of Combination of Individual Modal Responses According to the Revised RG1.92
Authors: Liu Shenghua , Xia Zufeng
Published: April 2011

The U.S. Nuclear Regulatory Commission (NRC) issued the revised regulatory guide 1.92 as Rev. 2 on July 2006, which is believed to efficiently reduce the unnecessary conservative which is introduced by combination methods specified in RG1.92 Rev.1 and to be acceptable for combining modal responses and spatial components in seismic response analysis of nuclear power plant structures, systems, and components (SSCs). But it is still impossible to use this method to do the combination of spectrum response analysis, because most popular general finite element programs do not develop a reasonable command or macros according to RG1.92 Rev.2. In order to perform the combination of individual modal responses according to this new guide conveniently, this paper developed a macro file used the ANSYS Parametric Design Language (APDL) based on ANSYS program which is widely used in nuclear industry. This paper examined the macro file with an actual support pipe, and compared the results got from the new method with the results acquired from previously accepted method such as double sum method and grouped method. The comparison result shows that the revised combination method actually reduced the unnecessary conservative previously accepted by method in RG1.92, Rev.1, also the results prove that the macro file this paper established is reasonable.
Experimental Analyses by SIMMER-III on Debris-Bed Coolability and Metallic Fuel Freezing Behavior
Authors: Yamano Hidemasa , Tobita Yoshiharu
Published: April 2011

This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: debris-bed coolability and metallic fuel freezing behavior. Related experimental database were reviewed to choose suitable experiments. To analyze the debris-bed coolability, the ACRR-D10 in-pile experiments were selected. SIMMER-III well simulated the heat transfer mechanisms including conduction, boiling and channeling observed in the experiment. Metallic fuel may freeze onto the stainless steel (cladding or wrapper tube) together with eutectic formation during core disruption in a metallic-fueled reactor. The CAFE´-UT2 experiment carried out using pure UO2 melt to investigate such phenomena was selected for the experimental analysis. In spite of no eutectic formation model in the SIMMER-III code, the calculated fuel penetration behavior was in good agreement with the experimental data.
Burn-Up Calibration and Error Analysis of HTR-10 Spherical Fuel Elements by Gamma Spectroscopy
Authors: Xia Bing , Li Fu , Wu Zongxin
Published: April 2011

The HTR-10 is a pebble bed High-temperature gas-cooled reactor (HTGR) with a nominal thermal power of 10 MW. The on-line nondestructive burn-up (BU) measurement for the spherical fuel elements with TRISO coated fuel particles is an important issue in the fuel management of the HTR-10. The HTR-10 employs an HPGe gamma spectroscopy system to determine the BU values of the discharged fuel elements, and the accuracy of the BU measurement system is a crucial issue. The calibration requirement and the absolute calibration method are discussed and analyzed in this paper. In this work, the long-lived fission product 137Cs is demonstrated as the best BU indicator in the HTR-10, whereas the inner-calibrating source method using the ratio of the amounts of 134Cs to 137Cs is not suitable. The BU value of a fuel element is proportional to the intensity of 137Cs gamma-ray and independent on the irradiation history of the fuel element, except for a small correction term related to the radioactive decay of 137Cs. The efficiency of the measurement system is calibrated by using a 137Cs standard source embedded into the center of a graphite sphere with the same size and the same graphite material as the fuel element, which was located at the measurement position of the fuel sphere, to eliminate the possible systematic errors from the measurement system. The difference in geometry and self-absorption between the BU measurement and the calibration measurement is corrected by applying the MCNP simulation. The error sources of the calibration results are analyzed, in which the uncertainty of the source activity (∼5%) is found to dominate. The other major error sources include the fluctuation of the uranium loading in the fuel elements, the errors of the nuclear data, the counting errors of the detector, and the correction procedure for the detection efficiency. More than 1,900 fuel elements discharged from the HTR-10 have been measured by far, and the BU values of these elements spread in the range from 8.4±0.5 GWD/t to 20±1.0 GWD/t, where the relative errors of the burn-ups consist of the calibration errors about 5.2% and the counting errors around the order of 1%. The behavior of the measured burn-ups has been discussed qualitatively.
Two-Phase Flow Measurement Studies for the SPES3 Integral Test Facility for IRIS Reactor Simulation
Authors: Greco Matteo , Ferri Roberta , Achilli Andrea , Gandolfi Stefano , Congiu Cinzia , Cattadori Gustavo , Bianchi Fosco , Monti Stefano , Luce Alfredo , Bertani Cristina , Mosetto Annamaria , Dzodzo Milorad
Published: April 2011

The measurement of two-phase flow parameters has never been an easy task in the experimental thermal-hydraulics and the need of such measurements in the SPES3 facility has led to investigation of different possibilities and evaluation methods to determine mass flows and energies. This paper deals with the theoretical prediction of the two-phase mass flow rate by the development of a mathematical model for a spool piece, consisting of a drag disk, a turbine and a void fraction detector. Data obtained by simulation of DBAs in the SPES3 facility, with the RELAP5 thermal-hydraulic code, have provided the reference conditions for defining the main thermal-hydraulic parameter ranges and selecting a set of instruments potentially suitable to measure and derive the required quantities. The governing equation and the instrumentation output are defined for each device. Three different turbine models (Aya, Rouhani and volumetric) have been studied to understand which one better adapts to two-phase flow conditions and to investigate the best instrument combination. The mathematical model has been tested versus the RELAP5 results with a reverse process where calculated variables, like void fraction, quality and slip ratio, are given as input to a specifically developed program to get back the mass flow rate. The analytical results, verified versus the DVI break transient, well agree with the RELAP5 mass flow rate. Specific tests on proper experimental loops are required to verify the analytical studies.
An Experimental Study of Bubble Nucleation Frequency in Subcooled Flow Boiling
Authors: Wang Guodong
Published: April 2011

In this paper, a simultaneous visualization and measurement study have been carried out to investigate bubble nucleation frequency of water in micro-channel at various heat fluxes and mass fluxes. A single micro-channel with an identical rectangular cross-section having a hydraulic of 137 μm and a heating length of 30 mm was used in this experiment. It is shown that the frequency of bubble nucleation increased drastically with the increase of heat flux and was also strongly dependent on the mass flux. A dimensionless frequency of bubble nucleation was correlated in terms of the Boiling number. The predictions of bubble nucleation frequency in the microchannel are found in good agreement with experimental data with a MAE of 10.4%.
A New Method on Flux Mapping by Function Expansion
Authors: Zhong Bin , Wang Kan , Yu Ganglin
Published: April 2011

The core flux (power) distribution is very important to safe and economical operation of nuclear reactor. It can be obtained by many methods depending on the desired accuracy and execution time. For on-line core surveillance and regulation, we need to get the real-time flux distribution. If the true local parameters such as fuel temperature, coolant temperature and material density were known, the solution of the diffusion equation with instantaneous parameters could, in principle, provide the necessary spatial details. However, in reality, it is impossible to obtain the operational “readings” of these parameters for each fuel cell. The detector results at certain locations can be applied to improve the results of the only diffusion calculations by Flux Mapping methods. Function expansion method is employed to express the approximate real distribution by the combination of several Flux Mapping method results as the expansion basis functions. The Harmonics Synthesis Method (HSM) and Least-Square method are combined to get a new Flux Mapping method in this paper. The simulation results show that the new method can be used for Flux Mapping and get better results.
AP1000 Operation Procedures System and Development
Authors: Wen Fang
Published: April 2011

This paper makes a brief introduction on AP1000 operation procedure system, including procedure classification, function and composition. In addition, key points of work flow process and the advantages of AP1000 operation procedures are described, among which the application of CPS (computerized procedure system) on AP1000 operation area and human factor engineering are highlighted. CPS, as an advanced procedure system, which is relatively new to existing nuclear power plants in China, does not only have the function of electronic indication for procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. Moreover, based on current situation, this paper offers several suggestions on procedure development for Sanmen AP1000 nuclear power project, i.e. first, we can ensure the quality of operation procedures by preparing a precise writer’s guideline, a friendly-interfaced procedure template, an efficient work configuration and an appropriate schedule; then determine the way how we are going to use operation procedures in English version; finally realize CPS Chinesization and localization gradually by digesting and absorbing API 000 technology from Westinghouse Electric Company. This paper gives an intact and systematic discourse on AP1000 operation procedure system and its characteristics. Besides, the latter part of this paper focuses on development of AP1000 operation procedures for Sanmen nuclear power plant and it would be a worthwhile reference for newly-built AP1000 units in China.
Simulation of B9401 Test in the RD-14M Experimental Facility With CATHARE2
Authors: Mladin Mirea , Mladin Daniela
Published: April 2011

RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.
Comparative Study of Government Subsidization of U.S. Electrical Energy Sources
Authors: Sandquist Gary M. , Kunze Jay F.
Published: April 2011

Perception that U.S. government energy subsidies have favored nuclear energy at expense of renewables (hydroelectric, wind, solar, geothermal) is not supported by facts. Largest beneficiaries between 1950 and 2006 from federal energy subsidies have been oil and gas receiving more than half of all federal incentives. Primary subsidy for nuclear energy has been R&D. Evaluating the actual electrical energy produced resulting from government subsidy support shows that wind and solar have cost taxpayers 355mils/kWh, coal 1.53 mils/kWh, nuclear 3.8 mils/kWh and hydro at 5.88 mils/kWh. Average cost of U.S. electrical energy in 2006 was 91 mils/kWh so renewables were subsidized at four times the average cost of electricity. Subsidy for Solar Photovoltaic to produce 0.01% of U.S. electricity as of 2006 was $4.43/kWh.
Climate Change Considerations in Sustainable Development of Nuclear Power Plants in the United States
Authors: Wan Ping K. , Carson Alice C. , Chan Desmond W.
Published: April 2011

Nuclear power generation has become an increasingly attractive alternative in the global power market due to growing demand for electric power, increasing global competition for fossil fuels, concern over greenhouse gas emissions and their potential impact on climate change, and the desire for energy independence. Nuclear energy plays an integral role in providing carbon free energy for sustainable development of global electric power generation. Assuring the protection of people and the environmental is a prime consideration in the design, construction, and operation of nuclear power plants. Potential environmental and safety concerns must be carefully evaluated and addressed. In order to assure that the nuclear power plant designs are sufficiently robust, the U.S. Nuclear Regulatory Commission (USNRC) requires that applicants for early site permits (ESP) and construction/operating licenses (COL) identify the most severe natural phenomena historically reported for the site and surrounding area to ensure sufficient design margin exists, considering the limited accuracy, quantity, and time in which the associated data has been collected. Because these permits are valid for a period up to 40 years, the potential impacts of climate change on the severity of natural phenomena, as it relates to the design basis and nuclear safety and environmental impacts are of increasing interest. Although no conclusive evidence or consensus of opinion is available on the long-term climatic changes resulting from human or natural causes, the USNRC has requested that climate change forecasts be considered for their potential affecting the most severe natural phenomena. The specific subject areas of concern include: • Extreme temperature and extreme precipitation (liquid & frozen) statistics – review 100 years of data around the site versus a review of the previous 30 years of data. • Extreme wind/basic wind speed – review previous 100 years of tropical cyclone data (including hurricanes) in the site vicinity versus previous 30 years of data. • Tornado – review of frequency and intensity trends and forecasts. • Drought – review of water availability / water supply during drought conditions and drought of record. • Stagnation Potential – review of conditions that would result in restrictive dispersion of greenhouse gas emissions. This paper examines the challenges and constraints in identifying and developing appropriate design- and operating-basis site/regional meteorological conditions while accounting for potential climate change during preparation of an ESP and/or COL. Because there is no regulatory guidance or quantitative acceptance criteria currently available, the methodology used to address climate change in a recent issued ESP will be discussed as an example.
The Effect of Fabrication Process on Microstructure Evolution of N18 Alloy Sheets
Authors: Dai Xun , Liu Hong , Jiang Hongman , Wang Ying
Published: April 2011

The microstructure and texture evolution of N18 sheets are studied after a series of fabrication processes which contain hot rolling at 780°C, cold rolling with the total deformation of 10% and recrystallized annealing at 580°C for 1.5h. The results show that grain size and the density of low angle grain boundaries of the fabricated sheets increase with the increase of hot rolling pass, and there is a little difference of grains for cold rolling and recrystallized sheets that deformed grain could be observed on both of them, which indicates that dynamic recrystallization occurs during hot rolling, but the cold deformation was so limited that the sheets could not be fully recrystallized because of lack of recrystallization driving force. Second phase precipitated particles (SPPs) in as-received N18 sheets are smaller than that in the fabricated sheets. It may because that part of SPPs are dissolved during hot rolling which lead to the volume decrease and supersaturation elements separate out to form SPPs again or primary SPPs grow up during recrystallization which lead to volume extending. The N18 sheets mainly have (0001) texture before and after fabrication and the density of basal pole which was along normal orientation on rolling plane increases as rolling pass increasing but decreases after cold rolling and increase again after recrystallization. After fabrication, the value of fND increases for N18 sheets, but there were a little difference among different rolling pass. EBSD analysis shows that the grains of rolling sheets mainly display orientations with <10 1¯ 0>//RD and {0001} <10 1¯ 0> texture was stronger, and the grains of recrystallized sheets mainly display orientations with <11 2¯ 0>//RD made {0001} <11 2¯ 0> texture increasing.
What Opinions Do People Have Through the Understanding Concerning High-Level Radioactive Waste Disposal Project?
Authors: Kimura Hiroshi , Furukawa Masashi , Sugiyama Daisuke , Chida Taiji
Published: April 2011

In Japan, the implementation of the high-level radioactive waste (HLW) disposal is one of urgent issues in the situation that Japan will continue the use of nuclear power. But, the lay people may not have the sufficient amount of information and knowledge about HLW disposal to hold their opinions about this issue. In this research, in order to clarify what opinions they will have with enough information and knowledge, we had the face-to-face dialogues about the HLW disposal with 2 or 3 lay persons. The dialogues were conducted 11 times with different lay persons’ groups. In these dialogues, after the lay participants had a certain amount of knowledge about HLW disposal, they became to talk about their opinions to the HLW disposal program in Japan. These opinions included the doubt against the open solicitation to select the siting area in the HLW disposal program of Japan, the emotion like NIMBY, the indication of lack of public relations about HLW disposal, and so on.
Non-Linear Time Series Analysis of the Infuence of Rolling Motion on Natural Circulation System
Authors: Zhang Wenchao , Tan Sichao , Gao Puzhen
Published: April 2011

Two-phase natural circulation flow instability under rolling motion condition was studied experimentally and theoretically. Experimental data were analyzed with nonlinear time series analysis methods. The embedding dimension, correlation dimension and K2 entropy were determined based on phase space reconstruction theory and G-P method. The maximal Lyapunov exponent was calculated according to the methods of small data sets. The nonlinear features of the two phase flow instability under rolling motion were analyzed with the results of geometric invariants coupling with the experimental data. The results indicated that rolling motion strengthened the nonlinear characteristics of two phase flow instability. Some typical nonlinear phenomena such as period-doubling bifurcations and chaotic oscillations were found in different cases.
N,N,N′,N′-Tetra-Methyl-3-Oxy-Pentane-1,5-Diamide (TMPDA): A Promising Back Extractant for Ln(III) and Zr(IV)
Authors: Wei Meng , He Qian’ge , Liu Xuegang , Chen Jing
Published: April 2011

Water-soluble oxa-diamide ligand, N,N,N′,N′-tetra-methyl-3-oxy-pentane-1,5-diamid (TMPDA) has been synthesized and purified. Its crystal structure, melting point, decomposition temperature, solubilities in aqueous phase and organic phase, distribution ratio between aqueous and organic phase, etc. are reported. The effect of TMPDA concentration in aqueous phase and HNO3 concentration in the equilibrium aqueous phase on the extraction efficiency of La(III), Ce(III), Pr(III), Nd(III), Zr(IV), Fe(III), Y(III), Mo(VI), Ru(III) and Pd(II) by 30% TRPO/kerosene have been studied. The results indicate that TMPDA dissolve well in aqueous phase but almost insoluble in kerosene or 30%TRPO/kerosene in the bi-phase system. It can effectively reduce the extraction of Ln(III), Y(III) and Zr(IV) into 30%TRPO/kerosene at a moderate acid system (0.24mol/L∼0.27mol/L HNO3). TMPDA is a promising stripping agent for Ln(III), Y(III) and Zr(IV) from loaded TRPO.
Stability of Radiation Grafted Membranes in Electro-Electrodialysis of HIX Solution
Authors: Tanaka Nobuyuki , Yamaki Tetsuya , Asano Masaharu , Maekawas Yasunari , Onuki Kaoru , Hino Ryutaro
Published: April 2011

Japan Atomic Energy Agency has been conducting research and development on a thermochemical water-splitting cycle featuring iodine- and sulfur-compounds (called an IS process) as one of promising heat utilization systems of High Temperature Gas-Cooled Reactors. We have prepared polymer electrolyte membranes by the radiation-induced graft polymerization and cross-linking methods and then have investigated their applicability to electro-electrodialysis (EED) for concentrating HI in an HI-I2-H2O mixture. For practical applications, EED membranes are required to be stable in the severe environment of high-temperature strongly acidic solutions. We thus examined thermal, chemical and electrochemical stabilities of the radiation-grafted membranes under the conditions of the actual EED operation over 100 hours, while measuring the time evolution of a cell voltage and a change in the ion exchange capacity between the EED experiment. The results showed that chemical cross-linking in the graft chains could largely improve the membrane stability.
Thermal Aging Effect of CDSS on the Integrity of the PWR Primary Pipe With Damage Tolerance Method
Authors: Wang Zhao-Xi , Xue Fei , Shi Hui-Ji , Yu Wei-wei , Shu Guo-gang
Published: April 2011

The structure integrity of the primary pipe from the Pressure Water Reactor (PWR) will be one of the most serous problems for the life extension of the reactor for that the toughness of the Duplex Stainless Steel will reduce sharply for thermal aging effect for a long time service at 350°C high temperature. To evaluate the structure integrity of the primary pipe, the material properties, such as the basic mechanical properties, the Charpy impact energy and the fracture properties for different aging times are investigated by macro mechanical tests. The stress analysis of the primary pipe for internal pressure, thermal stress and static stress for weight are performed. With the damage tolerance method, the residual toughness of the material and primary pipe structures are investigated according to the stress analysis of the material and the damage mechanics. From the experimental results, it is found that the strength increases while the toughness decreases with the increment of the accelerated aging time. The safety factor of the structure strength increases while that of the toughness decreases.
Prediction of Irradiation Embrittlement for Chinese Domestic A508-3 Steel
Authors: Lu Feng , Wang Rongshan , Huang Ping , Qian Haiyang
Published: April 2011

Reactor Pressure Vessel (RPV) is one of the most important components in nuclear power plant (NPP). The aging mechanism of primary concern for RPV is irradiation embrittlement, which can result in a decrease of fracture toughness of RPV steel. Prediction of irradiation embrittlement for a certain Chinese domestic manufactured A508-3 steel is performed. The calculation results given by the US safety standards, the French RCC-M standards and other international safety standards are compared with each other and compared to the data from commercial operation NPP surveillance program. The effect of neutron fluence is also investigated. Furthermore, the property of the steel against irradiation embrittlement is evaluated with the regulatory requirements in the relevant standards. It can be predicted that the Chinese domestic steel satisfies the requirements in these standards.