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...00 Add to Cart Molten Carbonate Fuel Cells: Modeling , Analysis, Simulation, and Control by Kai Sundmacher (Editor), Achim Kienle (Editor), Hans Josef Pesch (Editor), Joachim F. Berndt (Editor), Gerhard Huppmann (Editor) ...

Science.gov (United States)

The invited paper and 11 research papers are included, which were presented at the 85th Special Coke Meeting, Coke Group, Nenryo kyokai opened in Oct., 1988. As the invited paper, ''Production and utilization of blast furnace coke in the future'' is print...

International Nuclear Information System (INIS)

Requirements on PSA for risk informed applications are expressed in different international documents. The ASME PRA standard published in spring 2002 is one such document, PSA requirements are also expressed in the European Utility Requirements (EUR) for new reactors. The Swedish PSA requirements are provided in the Swedish regulators (SKI) statutes SKIFS 2004:1. SKI also has a review handbook for PSA activities (SKI report 2003:48). The review handbook is a support during review of the utilities PSA activities and the PSAs themselves. The review handbook expresses SKIs expectations by providing so called important aspects for both the PSA work and the PSAs, A comparison of SKIFS requirements and the important aspects in the Review handbook, on one side, and the requirements on PSA in EUR and ASME on the other side, is presented. The comparison shows a large difference in the level of detail in the different documents, where ...

Energy Technology Data Exchange (ETDEWEB)

For the purpose of digging up themes of the joint research which develop the R and D in the industrial technology field in Japan to a new stage, researchers were sent to the world representing research institutes to conduct the research survey of 'Nano-structured carbon and hydrogen absorption' and 'Development of the creation technology of nano-porous materials.' As to the former, an experiment on electrochemical hydrogen absorption of carbon materials including nanotubes was conducted by researchers dispatched, but the large absorption amount was not observed. As to the latter, visits were paid to Fraunhofer Institute and the related facilities in Germany, Princeton University, MIT, GIT and Naval Research Laboratories in the U.S., Orleans University in France, AO Research Institute (bone repair study) in Switzerland, Cambridge University and University of Bristol in the U.K., etc., and the research survey was made on the technology for making ...

Energy Technology Data Exchange (ETDEWEB)

ASME Subcommittee 2 on materials presented a series of questions to PVRC regarding the acceptability of using the criteria of ASME Section 2, Part D, Appendix 1 for extending design limits for AISI 304 stainless steel beyond 1,500 F to 2,000 F and Alloy 800 HT from 1,650 F to 1,800 F respectively. This paper describes a project supported by PVRC to try and find an answer to this question. The project consisted of three parts. The first was a survey to determine the intent behind the wording of the ASME criteria in order to make an extrapolation of methods for setting design limits to higher temperatures. The second was a demonstration of a methodology for developing very high temperature limits, using a set of creep data for Alloy 800 HT. The third was a parametric study to evaluate the feasibility of using the minimum creep rate based deformation criterion used in the ASME Code to set strain related ...

Energy Technology Data Exchange (ETDEWEB)

In 1998 the American Society of Mechanical Engineers (ASME) formed the Committee on Nuclear Risk Management (CNRM) and a Project Team to develop a standard on PRAs for use in risk informed applications. This ASME standard is being developed to help provide an adequate level of quality in PRAs that are being used to support ASME initiatives to risk informed in-service inspection (ISI) and in-service testing (IST) of nuclear power plant components. A related need supported by the industry and the U.S. Nuclear Regulatory Commission is to reduce the level of effort that is being expended in pilot applications of risk informed initiatives to address questions about the sufficiency of quality in the supporting PRA models. The purpose of this paper is to discuss the authors' views on some of the technical issues that were encountered in the effort to develop the ASME PRA standard. Draft 12 of this ...

Energy Technology Data Exchange (ETDEWEB)

The current Risk-informed Regulation (RIR) framework employs both a Level 1 Core Damage Frequency (CDF) and a Level 2 Large Early Release Frequency (LERF) as two surrogate measures for the plant risk. For their use in making regulatory decisions for the plant risk, it has been required to maintain an appropriate level of quality for the plant risk models. The first step for this purpose is to evaluate the quality of the PSA model in a qualitative or quantitative manner. Recently, a similar type of quality evaluation has been made for the KSNP Level 1 PSA model based on an ASME PRA Standard for the RIR application, but not for the corresponding LERF model. The main objective of this paper is to provide the ASME PRA Standard-based evaluation result for the existing KSNP Level 2 LERF model and the insights obtained from the evaluation process.

Science.gov (United States)

... Fuels", Paper AIAA-91-2451, 27th AIAA/SAE/ASME ... to-surface ballistic missiles than space-based space-to ... an effective hit-to-kill weapon without the ...

Science.gov (United States)

Gerard, G. and Papirno, R. , "Ductility Ratio of Aged Beta Titanium Alloy,. I t. Transactions of the ASM, VO~. 55, NO. 3, pp. 373-388, Sept. 1962. ...

Energy Technology Data Exchange (ETDEWEB)

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage ...

Energy Technology Data Exchange (ETDEWEB)

Dynamic fracture toughness tests were performed on materials which had been subjected to one of three long-time post weld type stress relief heat treatments: 48 hours at 1000/degree/F (538/degree/C), 24 hours at 1125/degree/F (607/degree/C), and 48 hours at 1125/degree/F (607/degree/C). Linear elastic K/sub Id/ results were obtained at low temperatures while J-integral techniques were utilized to evaluate dynamic fracture toughness over the transition and upper shelf temperature ranges. Tensile, Charpy impact, and drop weight nil-ductility transition tests as well as room temperature, air environment fatigue crack growth rate tests (SA508 Cl 2a only) were also performed. The fracture toughness of both materials exceeded the ASME specified minimum reference toughness K/sub IR/ curve. 17 refs.

Energy Technology Data Exchange (ETDEWEB)

In the RIR (Risk-informed Regulation), PSA (Probabilistic Safety Assessment) plays a major role because it provides overall risk insights for the regulatory body and utility. Therefore, the scope, the level of details and the technical adequacy of PSA, i.e. the quality of PSA is to be ensured for the successful RIR. To improve the quality of Korean PSA, we evaluate the quality of the KSNP (Korean Standard Nuclear Power Plant) internal full-power PSA model based on the 'ASME PRA Standard' and the 'NEI PRA Peer Review Process Guidance.' As a working group, PSA experts of the regulatory body and industry also participated in the evaluation process. It is finally judged that the overall quality of the KSNP PSA is between the ASME Standard Capability Category I and II. We also derive some items to be improved for upgrading the quality of the PSA up to the ASME Standard Capability Category II. ...

Energy Technology Data Exchange (ETDEWEB)

R and D of the robot control system was conducted in the following items: 1) integrated open control system, 2) remote control robot manipulation language, 3) human factor robot use built-in LAN system, 4) built-in actuator driver. In 1), there were some problems to be pointed out around the system, but the effectiveness was confirmed as system architecture of each verification item. In 2), development/design were made of RCML(R-Cube Manipulation Language) as a remote robot manipulation language, telecommunication protocol, and the experimental system, and the international standardization was targeted. In 3), the R and D was conducted of the realtime telecommunication protocol which clears the standards for the distributed control required for construction of human factor robot and the advanced realtime micro-controller, ULSI, which is the one that the protocol was made IC. In 4), an intelligent connector for built-in actuator was developed which enables saving of wiring in robot ...

Science.gov (United States)

In August 1995, the Nuclear Regulatory Commission (NRC) issued a policy statement proposing improved regulatory decisionmaking by increasing the use of PRA [probabilistic risk assessment] in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data. A key aspect in using PRA in risk-informed regulatory activities is establishing the appropriate scope and attributes of the PRA. In this regard, ASME decided to develop a consensus PRA Standard. The objective is to develop a PRA Standard such that the technical quality of nuclear plant PRAs will be sufficient to support risk-informed regulatory applications. This paper presents examples recommendations for the systems analysis element of a PRA for incorporation into the ASME PRA Standard.

Energy Technology Data Exchange (ETDEWEB)

Since the very high temperature reactor (VHTR) components such as hot gas ducts and intermediate heat exchanger (IHX) are operated during a design life of 30 to 60 years at 950 .deg. C and 3-8MPa in He impurities, their components are required to have good high temperature strength, creep-rupture strength, high temperature stability, and good corrosion resistance. Among them, creep properties are the most important, because the integrity of their components should be preserved during a design life of over 30 years at the maximum operating temperature up to 1000 .deg. C. So far, there are no materials approved by ASME III which is a nuclear code for temperatures reaching 1000 .deg. C, and until now, a material selection is still left as a key issue. Some leading candidate alloys can be selected from the high temperature materials approved by ASME VIII. Alloy 617 is considered as a prospective material for the IHX because it has an excellent ...

Science.gov (United States)

A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key requirement of the ...

Energy Technology Data Exchange (ETDEWEB)

The code of ASME Sec. III prescribes the general rules upon the design of a NSSS (nuclear steam supply system). The code provides further flexibility to the design of the nuclear structures by introducing a design by analysis concept. But it still preserves the conservatisms in design works by imposing strict failure mechanism and controlling material properties in use. A designer should prove the integrity of a structure under consideration by comparing the stress intensity, which was driven from the linearization of stress at concerning section, with the prescribed one. The recent development in computing system has enabled the commercial finite element programs to be a prevailing way to structural analysis field. But only few programs provide the procedure for stress linearization through the post-processing stage. Therefore, the simplified method which uses nodal stresses over the concerning section is introduced instead. But the issues related to the accuracy ...

Energy Technology Data Exchange (ETDEWEB)

In this study, the state-of-the art for estimating LERF is considered for the regulatory risk-informed decisions. The consideration is mainly focused on (1) the relationship between Level 2 PSA and LERF evaluation methodology, (2) the standard requirements in terms of modeling preparation and the acceptance criteria based on the application capability II of ASME PRA standard, and (3) some pending issues for developing and proposing a simplified LERF model. This study is preliminarily presented and will be updated for establishing detailed evaluation scheme of extended MPAS (multi-purpose probabilistic analysis of safety) model and preparing the technical basis.

Energy Technology Data Exchange (ETDEWEB)

This definitive reference is organized in an easy-to-follow format based on materials properties. It features new and existing data verified through major professional societies in the materials fields, such as ASM International and the American Ceramic Society. The third edition has been significantly expanded, most notably by the addition of new tabular material for a wide range of nonferrous alloys and various materials. The contents include: Structure of materials; Composition of materials; Phase diagram sources; Thermodynamic and kinetic data; Thermal properties of materials; Mechanical properties of materials; Electrical properties of materials; Optical properties of materials; Chemical properties of materials.

Energy Technology Data Exchange (ETDEWEB)

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA ...

Energy Technology Data Exchange (ETDEWEB)

As a part of the WE-NET project, the tanker for liquid hydrogen transport was studied. In fiscal 1996, some experiments and numerical analyses were proposed which are necessary to solve technological issues extracted in fiscal 1995 for heat insulation structure. The issue was roughly classified into vacuum and non-vacuum insulation, and their basic functions and required performance were arranged. Boil-off rate of 0.2-0.4%/d was targeted. The insulation system which applies polyurethane form (PUF) to tank surfaces and injects atmospheric N2 gas into the surrounding hold space, could achieve the targeted insulation performance by PUF of 1m in thickness. The system of vacuum panel insulation and atmospheric N2 gas injection into a hold space required the panel of 500mm in thickness because of the large effect of metallic outer panel material. The system of vacuum hold and PUF panels was faced with the essential issue for realizing and maintaining vacuum hold. The super insulation system ...

Energy Technology Data Exchange (ETDEWEB)

The feasibility study was conducted on a model project in Vietnam, aimed at solving the environmental pollution problems resulting from use of coal by demonstrating and disseminating the Japan's environmental technologies in the Southeast Asian countries. The feasibility study was conducted for the Cua Ong Coal Preparation Enterprise, which has the largest coal preparation capacity in Vietnam and port facilities. It is treating raw coal from 10 coal mines for classification and preparation, and shipping coal of various types that meet the standards for domestic use and export. The survey results point out that unrecovered coal remains in waste water discharged from the coal preparation plants to pollute the sea area, and that quantity of the refuse increases because of the unrecovered coal it contains. The environmental technologies needed to introduce include modification to variable wave pattern type jigging separator, refuse height measuring instrument and automatic ...

International Nuclear Information System (INIS)

This standard provides requirements and guidelines for installation, inspection and testing activities that assure the quality of important mechanical parts of a nuclear power plant not covered by the ASME Boiler and Pressure Vessel Code, Section III, during construction. These parts include those mechanical systems and components whose satisfactory performance is required: for the plant to operate reliably; to prevent accidents that could cause undue risk to the health and safety of the public; or to mitigate the consequences of such accidents if they were to occur. The requirements of this standard deal with the protection and control necessary to assure that the requisite quality of those important parts of the plant are preserved from the time items are removed from storage or receiving until they are incorporated into the plant up to but not including fuel loading for PWR plants and the completion of cold functional testing for BWR and HTGR plants.

Energy Technology Data Exchange (ETDEWEB)

The needs for more reasonable estimations for rare and extremely rare initiating events (IEs) have been reported in US peer review results. The American Society of Mechanical Engineers (ASME) PRA standard also proposes guidelines and requirements about the issues. Recently, US NRC addressed problems and the conservative assumptions on loss-of-coolant accident (LOCA) analysis and attempted to establish more rigorous methodology for estimating the frequencies depending on break size. The results of peer reviews for KHNP reference plants also represented that the data used in estimating IEs were outdated and the methodology also needed to be improved. In this paper, for more appropriate estimation of rare and extremely rare initiating events (IEs), e.g., LOCAs and steam generator tube ruptures (SGTRs), a new approach considering expert elicitation process is presented and corresponding core damage frequency (CDF) is calculated

British Library Electronic Table of Contents (United Kingdom)

This paper describes a comprehensive model of wastewater treatment in secondary facultative ponds, which combines 3D hydrodynamics with a mechanistic water quality model. The hydrodynamics are based on the Navier-Stokes equation for incompressible fluids under shallow water and Boussinesq assumptions capturing the flow dynamics along length, breadth and depth of the pond. The water quality sub model is based on the Activated Sludge Model (ASM) concept, describing COD and nutrient removal as function of bacterial growth following Monod kinetics, except for Escherichia coli removal, which was modelled as first order decay. The model was implemented in the Delft3D software and was used to evaluate the effect of wind and the addition of baffles on the water flow pattern, temperature profiles i...

International Nuclear Information System (INIS)

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be ...

International Nuclear Information System (INIS)

Since 1978, an important effort has been on-going between Framatome and Electricite de France (EDF) to develop and update technical rules for the design and construction of nuclear power plants. Two design codes are developed, one for PWRs: RCC-M and one for FBRs: RCC-MR. After some comments on the organization of RCCM Commission, the paper presents: (1) the major differences with the ASME Section III Code for fatigue, plastic shakedown and rupture analysis; (2) the review of some important facts in terms of availability or maintenance costs (in-service inspection, replacements, repairs, etc.) of operating plants like underclad cracking, stratification, CRDM nozzles, cast elbows, steam generator tubes, etc.; (3) how can one take into consideration this operating feedback in RCC-M Code; and (4) few examples of major on-going developments for fracture analysis on different components and fatigue analysis of valves. In conclusion, the standardization of the plants is ...

Energy Technology Data Exchange (ETDEWEB)

The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. We then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. We concluded that these changes substantially reduce calculated piping responses and allows piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the ...

Energy Technology Data Exchange (ETDEWEB)

A preliminary guideline for the design and evaluation of LMR high temperature structure is presented based upon ASME B and PV Code, Section III, Subsection NH. The main contents of this guideline are the materials, general design, vessel, piping, core support structure, pumps, valves, fabrication, examination, and testing for the class 1 components. The ratcheting evaluation, enhanced creep assessment, welds design and evaluation, inelastic analysis approach, piping design alternatives, and bellows design method are described in appendices. A user of this guideline should follow the essential procedures and may refer to other pertinent codes, standards, laws, regulations, or other pertinent documents when this guideline does not lead to proper design of the structure. While this guideline adopts major procedures of Subsection NH, it refers to the RCC-MR and/or DDS in some amount for the items where these codes have excellency to improve this guideline.

Energy Technology Data Exchange (ETDEWEB)

This paper describes the activities for improving the Probabilistic Safety Assessment (PSA) quality in the human reliability analysis (HRA) of the pre-accident human errors for the Korea Standard Nuclear Power Plant (KSNP). We evaluate the HRA results of the PSA for the KSNP and identify the items to be improved using the ASME PRA Standard. Evaluation results show that the ratio of items to be improved for pre-accident human errors is relatively high when compared with the ratio of those for post-accident human errors. They also show that more than 50% of the items to be improved for pre-accident human errors are related to the identification and screening analysis for them. In this paper, we develop the modeling guidelines for pre-accident human errors and apply them to the auxiliary feedwater system of the KSNP. Application results show that more than 50% of the items to be improved for the pre-accident human errors of the auxiliary feedwater system are resolved. ...

International Nuclear Information System (INIS)

The quality assurance requirements in use today for radioactive waste facilities, geologic repositories and hazardous waste projects were developed initially for the nuclear power plant industry, and their intent is being applied by regulations and guidance documents to radioactive and hazardous waste programs. The wording of the NRC quality assurance requirements in Appendix B of 10CFR50, the related guidance documents and the industry's ANSI/ASME NQA-1 were developed over a period of several years to address quality assurance for the design and construction of the complex and interactive systems to produce electrical power using nuclear fuel. Now, those same documents are the basis for the quality assurance requirements and guidance for waste management facilities and repositories. The intent of Appendix B of 10CFR50 and NQA-1 can easily be applied to waste projects, providing one understands and uses the intent of the requirements. This paper describes the ...

Energy Technology Data Exchange (ETDEWEB)

For a reliable design against fatigue failure of high-temperature components of different types of the heat resistant material Alloy 800, verified parameters of the low-cycle fatigue (LCF) behaviour are required. These parameters were obtained by a scatterband evaluation of the results of low-cycle fatigue tests, using a computerized data bank which includes all available actual data sets. The results of the evaluation are presented as mean fatigue curves with the upper and lower boundary limits of the scatterbands. For the type Alloy 800 in the recrystallization annealed condition, the results cover a temperature range from 20 to 750deg C; for the solution annealed types they cover a temperature range from 20 to 950deg C. Specimens tested in HTR helium at high temperatures had longer lives than specimens tested in air. This effect was represented quantitatively by the ratio of the total strain ranges from tests in HTR helium and air as a function of test temperature and cycles to ...

International Nuclear Information System (INIS)

During the last ten years, the U.S. Nuclear Regulatory Commission and the U.S. nuclear utilities have been developing methods and requirements for risk-informed applications making use of probabilistic risk assessments (PRA) of nuclear power plants. Early in this process, it became clear that the existing PRAs were done with different objectives and methodologies by different analysts. For uniformity and consistency in future risk-informed applications, industry consensus standards on probabilistic risk assessments were deemed to be essential. Currently, the following standards have been published or under preparation: - ASME RA-S-2002: 'Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications', Addendum C, March 2005. - ANSI/ANS-58.21-2003 'External-Events PRA Methodology' March 2003. - ANS-58.22 'Low Power and Shutdown PRA Standard'. - ANS-58.23 'Fire PRA Methodology Standard'. - ANS Level 2 and Level 3 PRA Standards. The ...

Energy Technology Data Exchange (ETDEWEB)

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-informed In-service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA ...

Energy Technology Data Exchange (ETDEWEB)

This study identifies potential applications and design requirements for ceramic materials (CMs) and ceramic composite materials (CCMs) in the NGNP hightemperature gas-cooled reactor (HTGR) primary circuit. Components anticipated for fabrication from non-graphite CMs and CCMs are identified along with recommended normal and off-normal operating conditions. The evaluation defines required dimensions and material properties of the candidate materials for normal operating conditions (NOC), anticipated transients, abnormal events, and design basis events. The report also identifies additional activities required for codifying the selected materials. The activities include ASTM Standard and ASME Code development and other work to support NRC licensing of the plant. Evaluation of the NGNP baseline design indicates components requiring either CMs or CCMs depend upon the reactor operating temperatures. For a reactor outlet temperature of 900 oC, four of the five evaluated ...

International Nuclear Information System (INIS)

During the service in nuclear power plants, steam generator (SG) tubing will suffer from a vibration with a frequency from 30-40 Hz. This means that SG tubing will undergo a very high cycle fatigue process during the design life time (25 years=about 3x10"1"0cycles). The material degradation and the very high cycle fatigue life of SG tubing materials due to the vibration are therefore of concern. In this investigation, the fatigue properties of Alloy 690 and Alloy 800 SG tubing in the very high cycle fatigue regime (higher than 10"7 cycles) and strain controlled fatigue at temperatures up to 330"oC have been investigated. The fatigue properties of SG tubing of Alloy 690 are comparable to that of wrought material. The strain controlled fatigue properties of Alloy 690 and Alloy 800 are slightly better at 288"oC than at room temperature (RT). At the stress near the fatigue endurance, the influence of temperature on the high cycle fatigue properties of Alloy 690 is small. The fatigue ...

International Nuclear Information System (INIS)

Up to the present, the generic values of the Common cause failure (CCF) event parameters have been used in most PSA projects for the Korean NPPs. However, the CCF analysis should be performed with plant specific information to meet Category II of the ASME PRA Standard. Therefore, we estimated the Alpha factor parameters of the emergency diesel generator (EDG) for Ulchin Unit 3 by using the International Common-Cause Failure data Exchange (ICDE) database. The ICDE database provides the member countries with only the information needed for an estimation of the CCF parameters. The Ulchin Unit A3, pressurized water reactor, has two onsite EDGs and one alternate AC (AAC) diesel generator. The onsite EDGs of Unit 3 and 4 and the AAC are manufactured by the same company, but they are designed differently. The estimation procedure of the Alpha factor used in this study follows the approach of the NUREG/CR-5485. Since we did not find any qualitative difference between the ...

Energy Technology Data Exchange (ETDEWEB)

Two mechanical design requirements are defined for the SSC Collider beam tube. First, the vacuum requirement (luminosity lifetime = 150 hrs). It requires the design of a pressure boundary within the cold mass vessel to provide a vacuum tunnel for the proton beam and to minimize the synchrotron radiation gas desorbtion with a suitable material. The Collider beam tube design is under an intensive activity to search for a material that will meet the luminosity requirement without a distributed pump or liner. Second is the tube wall`s resistivity requirement ({sigma}*t = 2E5 {Omega}{sup {minus}1}). For a 4.2 K beam tube the Cu thickness is 100 {mu}m (RRR=30,6.7 T, {sigma}=2E9{Omega}{sup {minus}1}m{sup {minus}1}). The copper yield strength is relatively low in comparison to steel and, therefore, the design of the steel layer is governed by the copper layer yield stress limit. A beam tube subjected to eddy current load in a quenching dipole requires an optimum diameter design to provide ...

International Nuclear Information System (INIS)

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the 'ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using ...

Energy Technology Data Exchange (ETDEWEB)

Human Reliability Analysis (HRA) is a very important part of Probabilistic Risk Analysis (PRA), and constant work is dedicated to improving methods, guidance and data in order to approach realism in the results as well as looking for ways to use these to reduce accident frequency at plants. Further, in order to advance in these areas, several HRA studies are being performed globally. Mexico has participated in the International HRA Empirical study with the objective of -benchmarking- HRA methods by comparing HRA predictions to actual crew performance in a simulator, as well as in the empirical study on a US nuclear power plant currently in progress. The focus of the first study was the development of an understanding of how methods are applied by various analysts, and characterize the methods for their capability to guide the analysts to identify potential human failures, and associated causes and performance shaping factors. The HRA benchmarking study has been performed by using the ...

International Nuclear Information System (INIS)

This paper identifies the event sequences that require thermal-hydraulic analyses for the success criteria of probabilistic safety analysis (PSA). The selection of the sequences is performed based on the review of the NEI Peer Review Process Guidance and ASME PRA Standard. Success criteria are the important element of PSA quality. Success criteria decide the success or failure of the key function in the PSA event tree. It is defined as a minimum set of components/trains of system required to mitigate an accident. Thermal-hydraulic codes are generally used to derive time-related criteria in the PSA, such as operator action time used in human reliability analysis (HRA), event timing, and time to recover the component or the power. This paper suggests the use of the MARS code for the T-H analysis to obtain the success criteria and sequence timing, and operator action time. In the Kori Units 3 and 4 PSA report, the T-H analyses for those criteria were performed by the ...

Energy Technology Data Exchange (ETDEWEB)

An investigation was conducted into alternative techniques for zinc and electrolyte regeneration and reuse in the refuelable zinc/air battery that was developed by LLNL and previously tested on a moving electric bus using cut wire. Mossy zinc was electrodeposited onto a bipolar array of inclined Ni plates with an energy consumption of 1.8 kWh/kg. Using a H{sub 2}-depolarized anode, zinc was deposited at 0.6 V (0.8 kA/m{sup 2}); the open circuit voltage was 0.45 V. Three types of fuel pellets were tested and compared with results for 0.75 mm cut wire: spheres produced in a spouted bed (UCB); coarse powder produced by gas-atomization (Noranda); and irregular pellets produced by chopping 1-mm plates of compacted zinc fines (Eagle-Picher, Inc.). All three types transported within the cell. The coarse powder fed continuously from hopper to cell, as did the compacted pellets (< 0.83 mm). Large particles (> 0.83 mm; Eagle-Picher and UCB) failed to feed from hopper into cell, ...