Application of NASA Kennedy Space Center system assurance analysis methodology to nuclear power plant systems designs

International Nuclear Information System (INIS)

Page, D.W.

1985-01-01

The Kennedy Space Center (KSC) entered into an agreement with the Nuclear Regulatory Commission (NRC) to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. In joint meetings of KSC and Duke Power personnel, an agreement was made to select to CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set a Final Safety Analysis Reports as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application. The conclusion is drawn that nuclear power plant systems and aerospace ground support systems are similar in complexity and design and share common safety and reliability goals. The SAA methodology is readily adaptable to nuclear power plant designs because of it's practical application of existing and well known safety and reliability analytical techniques tied to an effective management information system

Design and analysis of nuclear processes with the APROS

International Nuclear Information System (INIS)

Haenninen, M.; Puska, E.K.; Nystroem, P.

1987-01-01

APROS (Advanced Process Simulator) is the product being developed in the process simulators project of Imatran Voima Co. and Technical Research Centre of Finland. The aim is to design and construct an efficient and easy to use computer simulation system for process and automation system design, evaluation, analysis, testing and training purposes. As halfway of this project a working system exists with a large number of proven routines and models. However, a lot of development is still foreseen before the project will be finished. This article gives an overview of the APROS in general and of the nuclear features in particular. The calculational capabilities of the system are presented with the help of one example. (orig.)

Advanced design nuclear power plants: Competitive, economical electricity. An analysis of the cost of electricity from coal, gas and nuclear power plants

International Nuclear Information System (INIS)

1992-06-01

This report presents an updated analysis of the projected cost of electricity from new baseload power plants beginning operation around the year 2000. Included in the study are: (1) advanced-design, standardized nuclear power plants; (2) low emissions coal-fired power plants; (3) gasified coal-fired power plants; and (4) natural gas-fired power plants. This analysis shows that electricity from advanced-design, standardized nuclear power plants will be economically competitive with all other baseload electric generating system alternatives. This does not mean that any one source of electric power is always preferable to another. Rather, what this analysis indicates is that, as utilities and others begin planning for future baseload power plants, advanced-design nuclear plants should be considered an economically viable option to be included in their detailed studies of alternatives. Even with aggressive and successful conservation, efficiency and demand-side management programs, some new baseload electric supply will be needed during the 1990s and into the future. The baseload generating plants required in the 1990s are currently being designed and constructed. For those required shortly after 2000, the planning and alternatives assessment process must start now. It takes up to ten years to plan, design, license and construct a new coal-fired or nuclear fueled baseload electric generating plant and about six years for a natural gas-fired plant. This study indicates that for 600-megawatt blocks of capacity, advanced-design nuclear plants could supply electricity at an average of 4.5 cents per kilowatt-hour versus 4.8 cents per kilowatt-hour for an advanced pulverized-coal plant, 5.0 cents per kilowatt-hour for a gasified-coal combined cycle plant, and 4.3 cents per kilowatt-hour for a gas-fired combined cycle combustion turbine plant

Interaction between core analysis methodology and nuclear design: some PWR examples

International Nuclear Information System (INIS)

Rothleder, B.M.; Eich, W.J.

1982-01-01

The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

Nuclear data for fission reactor core design and safety analysis: Requirements and status of accuracy of nuclear data

International Nuclear Information System (INIS)

Rowlands, J.L.

1984-01-01

The types of nuclear data required for fission reactor design and safety analysis, and the ways in which the data are represented and approximated for use in reactor calculations, are summarised first. The relative importance of different items of nuclear data in the prediction of reactor parameters is described and ways of investigating the accuracy of these data by evaluating related integral measurements are discussed. The use of sensitivity analysis, together with estimates of the uncertainties in nuclear data and relevant integral measurements, in assessing the accuracy of prediction of reactor parameters is described. The inverse procedure for deciding nuclear data requirements from the target accuracies for prediction of reactor parameters follows on from this. The need for assessments of the uncertainties in nuclear data evaluations and the form of the uncertainty information is discussed. The status of the accuracies of predictions and nuclear data requirements are then summarised. The reactor parameters considered include: (a) Criticality conditions, conversion and burn-up effects. (b) Energy production and deposition, decay heating, irradiation damage, dosimetry and induced radioactivity. (c) Kinetics characteristics and control, including temperature, power and coolant density coefficients, delayed neutrons and control absorbers. (author)

Failure modes and effects analysis as a design tool for nuclear safety systems

International Nuclear Information System (INIS)

Tashjian, B.M.

1975-01-01

The activities of nuclear power plant designers are monitored by government and industry to an unprecedented degree. This involves not only rigid design and quality assurance criteria, but extensive documentation and reporting. The failure modes and effects analysis (FMEA) is a technique for checking designs and assuring quality. Included in the FMEA is a system of documentation. A simplified example of the reactor protective system (RPS) is used to illustrate the method. (U.S.)

Design requirements for new nuclear reactor facilities in Canada

International Nuclear Information System (INIS)

Shim, S.; Ohn, M.; Harwood, C.

2012-01-01

The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

The analysis of FCA critical experiments and its application to ''JOYO'' nuclear design

International Nuclear Information System (INIS)

Iijima, S.

1979-01-01

A series of extensive mockup experiments in support of Japanese Experimental Fast Reactor, ''JOYO'', were performed at Fast Critical Assembly in JAERI, from February 1970 to March 1972. The present paper describes the results of analysis of these mockup experiments and its application to ''JOYO'' nuclear design. The basic calculational method of the analysis is the same as that employed in ''JOYO'' neutronics calculation, viz., the 6-group diffusion theory using 25-group NAIG Nuclear Set No. 5. Corrections to the base calculations were evaluated by using one-dimensional S 4 transport theory and integral transport theory. The ABBN group constants were also used for the sake of comparison. The most probable values of JOYO neutronics parameters were determined by applying the bias factor (E/C) to the calculated values. The uncertainties of the most probable values were also determined, and they were taken into consideration in the JOYO design

Research on digital system design of nuclear power valve

Science.gov (United States)

Zhang, Xiaolong; Li, Yuan; Wang, Tao; Dai, Ye

2018-04-01

With the progress of China's nuclear power industry, nuclear power plant valve products is in a period of rapid development, high performance, low cost, short cycle of design requirements for nuclear power valve is proposed, so there is an urgent need for advanced digital design method and integrated design platform to provide technical support. Especially in the background of the nuclear power plant leakage in Japan, it is more practical to improve the design capability and product performance of the nuclear power valve. The finite element numerical analysis is a common and effective method for the development of nuclear power valves. Nuclear power valve has high safety, complexity of valve chamber and nonlinearity of seal joint surface. Therefore, it is urgent to establish accurate prediction models for earthquake prediction and seal failure to meet engineering accuracy and calculation conditions. In this paper, a general method of finite element modeling for nuclear power valve assembly and key components is presented, aiming at revealing the characteristics and rules of finite element modeling of nuclear power valves, and putting forward aprecision control strategy for finite element models for nuclear power valve characteristics analysis.

Seismic design ampersand analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid open-quotes deterministicclose quotes and open-quotes probabilisticclose quotes concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Seismic design and analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid ''deterministic'' and ''probabilistic'' concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Use of risk assessment methods for security design and analysis of nuclear and radioactive facilities

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Andrade, Marcos C.; Jordao, Elizabete

2011-01-01

The objective of this work is to evaluate the applicability of risk assessment methods for analyzing the physical protection of nuclear and radioactive facilities. One of the important processes for physical protection in nuclear and radioactive facilities is the identifying of areas containing nuclear materials, structures, systems or components to be protected from sabotage, which could directly or indirectly lead to unacceptable radiological consequences. A survey of the international guidelines and recommendations about vital area identification, design basis threat (DBT), and the security of nuclear and radioactive facilities was carried out. The traditional methods used for quantitative risk assessment, like FMEA (Failure Mode and Effect Analysis), Event and Decision Trees, Fault and Success Trees, Vulnerability Assessment, Monte Carlo Simulation, Probabilistic Safety Assessment, Scenario Analysis, and Game Theory, among others, are highlighted. The applicability of such techniques to security issues, their pros and cons, the general resources needed to implement them, as data or support software, are analyzed. Finally, an approach to security design and analysis, beginning with a qualitative and preliminary examination to determine the range of possible scenarios, outcomes, and the systems to be included in the analyses, and proceeding to a progressively use of more quantitative techniques is presented. (author)

Nuclear power. Volume 1. Nuclear power plant design

International Nuclear Information System (INIS)

Pedersen, E.S.

1978-01-01

NUCLEAR POWER PLANT DESIGN is intended to be used as a working reference book for management, engineers and designers, and as a graduate-level text for engineering students. The book is designed to combine theory with practical nuclear power engineering and design experience, and to give the reader an up-to-date view of the status of nuclear power and a basic understanding of how nuclear power plants function. Volume 1 contains the following chapters; (1) nuclear reactor theory; (2) nuclear reactor design; (3) types of nuclear power plants; (4) licensing requirements; (5) shielding and personnel exposure; (6) containment and structural design; (7) main steam and turbine cycles; (8) plant electrical system; (9) plant instrumentation and control systems; (10) radioactive waste disposal (waste management) and (11) conclusion

AZTLAN platform: Mexican platform for analysis and design of nuclear reactors

International Nuclear Information System (INIS)

Gomez T, A. M.; Puente E, F.; Del Valle G, E.; Francois L, J. L.; Martin del Campo M, C.; Espinosa P, G.

2014-10-01

The Aztlan platform Project is a national initiative led by the Instituto Nacional de Investigaciones Nucleares (ININ) which brings together the main public houses of higher studies in Mexico, such as: Instituto Politecnico Nacional, Universidad Nacional Autonoma de Mexico and Universidad Autonoma Metropolitana in an effort to take a significant step toward the calculation autonomy and analysis that seeks to place Mexico in the medium term in a competitive international level on software issues for analysis of nuclear reactors. This project aims to modernize, improve and integrate the neutron, thermal-hydraulic and thermo-mechanical codes, developed in Mexican institutions, within an integrated platform, developed and maintained by Mexican experts to benefit from the same institutions. This project is financed by the mixed fund SENER-CONACYT of Energy Sustain ability, and aims to strengthen substantially to research institutions, such as educational institutions contributing to the formation of highly qualified human resources in the area of analysis and design of nuclear reactors. As innovative part the project includes the creation of a user group, made up of members of the project institutions as well as the Comision Nacional de Seguridad Nuclear y Salvaguardias, Central Nucleoelectrica de Laguna Verde (CNLV), Secretaria de Energia (Mexico) and Karlsruhe Institute of Technology (Germany) among others. This user group will be responsible for using the software and provide feedback to the development equipment in order that progress meets the needs of the regulator and industry; in this case the CNLV. Finally, in order to bridge the gap between similar developments globally, they will make use of the latest super computing technology to speed up calculation times. This work intends to present to national nuclear community the project, so a description of the proposed methodology is given, as well as the goals and objectives to be pursued for the development of the

Application of NASA Kennedy Space Center System Assurance Analysis methodology to nuclear power plant systems designs

International Nuclear Information System (INIS)

Page, D.W.

1985-01-01

In May of 1982, the Kennedy Space Center (KSC) entered into an agreement with the NRC to conduct a study to demonstrate the feasibility and practicality of applying the KSC System Assurance Analysis (SAA) methodology to nuclear power plant systems designs. North Carolina's Duke Power Company expressed an interest in the study and proposed the nuclear power facility at CATAWBA for the basis of the study. In joint meetings of KSC and Duke Power personnel, an agreement was made to select two CATAWBA systems, the Containment Spray System and the Residual Heat Removal System, for the analyses. Duke Power provided KSC with a full set of Final Safety Analysis Reports (FSAR) as well as schematics for the two systems. During Phase I of the study the reliability analyses of the SAA were performed. During Phase II the hazard analyses were performed. The final product of Phase II is a handbook for implementing the SAA methodology into nuclear power plant systems designs. The purpose of this paper is to describe the SAA methodology as it applies to nuclear power plant systems designs and to discuss the feasibility of its application. (orig./HP)

Nuclear analysis for ITER

International Nuclear Information System (INIS)

Santoro, R.T.; Iida, H.; Khripunov, V.; Petrizzi, L.; Sato, S.; Sawan, M.; Shatalov, G.; Schipakin, O.

2001-01-01

This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shut-down. (author)

Application of nuclear power station design criteria to non-nuclear installations

International Nuclear Information System (INIS)

Regan, J.D.; Hughes, D.J.

1989-01-01

The nuclear industry is multi faceted, in that it includes large and complex chemical plants, a large number of different types of nuclear power stations, and on shore ship maintenance facilities, each with its own unique problems. Since the early days the industry has been aware of the additional problem which is superimposed on what may be classed as traditional fire risks, that is, the risk of an uncontrolled release of radioactivity. This has led to the development of sophisticated fire prevention and control techniques which are applied to new plants, and to the backfitting of older plants. The techniques of analysis, design and operation can be applied to both nuclear and non-nuclear installations. Passive protection is preferred backed up by active techniques. Segregation of essential plant to increase the probability of sufficient surviving to ensure safety systems operate and the provision of smoke free, protected escape routes are important aspects of layout and design. Reliability assessments, venting of smoke and hot gases, fire severity analysis, application of mathematical models contribute to the final design to protect against fires. Experiences built up in the fire fighting profession is integrated into the numerical approach by frequent involvement of the local Fire Officers at each stage of the design and layout of installations. (author)

Seismic design of nuclear power plants - an assessment

International Nuclear Information System (INIS)

Howard, G.E.; Ibanez, P.; Smith, C.B.

1976-01-01

This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the 'nuclear power industry'. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation. (Auth.)

Optimum design of a nuclear heat supply

International Nuclear Information System (INIS)

Borel, J.P.

1984-01-01

This paper presents an economic analysis for the optimum design of a nuclear heat supply to a given district-heating network. First, a general description of the system is given, which includes a nuclear power plant, a heating power plant and a district-heating network. The heating power plant is fed with steam from the nuclear power plant. It is assumed that the heating network is already in operation and that the nuclear power plant was previously designed to supply electricity. Second, a technical definition of the heat production and transportation installations is given. The optimal power of these installations is examined. The main result is a relationship between the network capacity and the level of the nuclear heat supply as a substitute for oil under the best economic conditions. The analysis also presents information for choosing the best operating mode. Finally, the heating power plant is studied in more detail from the energy, technical and economic aspects. (author)

Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

International Nuclear Information System (INIS)

Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

1991-01-01

This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

Design of nuclear fuel cells by means of a statistical analysis and a sensibility study

International Nuclear Information System (INIS)

Jauregui C, V.; Castillo M, J. A.; Ortiz S, J. J.; Montes T, J. L.; Perusquia del C, R.

2013-10-01

This work contains the results of the statistical analysis realized to study the nuclear fuel cells performance, considering the frequencies for the election of fuel bars used in the design of the same ones. The election of the bars used for the cells design are of 3 types, the first election shows that to the plotting the respective frequency is similar to a normal distribution, in the second case the frequencies graph is of type inverted square X 2 and the last election is when the bars are chosen in aleatory form. The heuristic techniques used for the cells design were the neural networks, the ant colonies and a hybrid between the dispersed search and the trajectories re-linkage. To carry out the statistical analysis in the cells design were considered the local power peak factor and the neutron infinite multiplication factor (k∞) of this. On the other hand, the performance of the designed cells was analyzed when verifying the position of the bars containing gadolinium. The results show that is possible to design cells of nuclear fuel with a good performance, when considering the frequency of the bars used in their design. (Author)

Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Gregg, D.W.; Hockman, J.N.; Russell, E.W.; Stein, W.

1984-01-01

This report discusses the selection and analysis of conceptual waste package developed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for possible disposal of high-level nuclear waste at a candidate site at Yucca Mountain, Nevada. The design requirements that the waste package must conform to are listed, as are several desirable design considerations. Illustrations of the reference and alternative designs are shown. Four austenitic stainless steels (316L SS, 321 SS, 304L SS and Incoloy 825 high nickel alloy) have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and ecnonmic analyses supporting the selection of the conceptual waste package designs is included. Postclosure containment and release rates are not analyzed in this report

High degree utilization of computers for design of nuclear power plants

International Nuclear Information System (INIS)

Masui, Takao; Sawada, Takashi

1992-01-01

Nuclear power plants are the huge technology in which various technologies are compounded, and the high safety is demanded. Therefore, in the design of nuclear power plants, it is necessary to carry out the design by sufficiently grasping the behavior of the plants, and to confirm the safety by carrying out the accurate design evaluation supposing the various operational conditions, and as the indispensable tool for these analysis and evaluation, the most advanced computers in that age have been utilized. As to the utilization for the design, there are the fields of design, analysis and evaluation and another fields of the application to the support of design. Also in the field of the application to operation control, computers are utilized. The utilization of computers for the core design, hydrothermal design, core structure design, safety analysis and structural analysis of PWR plants, and for the nuclear design, safety analysis and heat flow analysis of FBR plants, the application to the support of design and the application to operation control are explained. (K.I.)

Applications of Axiomatic Design in Developing Nuclear Systems

Energy Technology Data Exchange (ETDEWEB)

Heo, Gyunyoung [Kyung Hee University, Seoul (Korea, Republic of)

2007-10-15

The first step of designing nuclear systems starts with the identification of the top-level requirements given by stake holders and regulatory authorities. A detailed design of structure, system and component then follows. Design is divided into two processes: 'synthesis' and 'analysis.' While synthesis is the process of decision making on parameters, analysis is the process of optimizing the parameters selected. It is known from experience that the mistakes made in the synthesis process, particularly of a conceptual stage, are never completely corrected in the analysis process, which is more serious in designing complex safety critical systems such as nuclear power plants. It should be also noted that we normally believe that synthesis is only driven by engineers' heuristic knowledge. This paper proposes the applications of Axiomatic Design (AD), which is a design management tool as slightly opposed to this conventional view. I hypothesize that the design management using design axioms reduces uncertainty and subjectivity particularly at a conceptual phase so that a safer nuclear system can be developed while reducing cost in view of the system's entire life cycle. I will describe the notion of AD and introduce a few case studies.

Analysis of effect of safety classification on DCS design in nuclear power plants

International Nuclear Information System (INIS)

Gou Guokai; Li Guomin; Wang Qunfeng

2011-01-01

By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants, especially the requirement of availability and reliability of I and C systems, the characteristics of modem DCS technology and I and C products currently applied in nuclear power field are interpreted. According to the requirements on the safety operation of nuclear power plants and the regulations for safety audit, the effect of different safety classifications on DCS design in nuclear power plants is analyzed, by considering the actual design process of different DCS solutions in the nuclear power plants under construction. (authors)

Meeting cross-section requirements for nuclear-energy design

Energy Technology Data Exchange (ETDEWEB)

Weisbin, C.R.; de Saussure, G.; Santoro, R.T. (Oak Ridge National Lab., TN (USA)); Gilai, T. (Ben-Gurion Univ. of the Negev, Beersheba (Israel))

1982-01-01

Current requirements in cross-section data that are essential to nuclear-energy programmes are summarized and explained and some insight into how these data might be obtained is provided. The six sections of the paper describe: design parameters and target accuracies; data collection, evaluation and analysis; determination of high-accuracy differential nuclear data for technological applications; status of selected evaluated nuclear data; analysis of benchmark testing; identification of important cross sections and inferred needs.

Considerations for Applying Design Extension Conditions to Domestic Nuclear Power Plants

International Nuclear Information System (INIS)

Ryu, Yongho

2013-01-01

The concept is designed to include more serious accidents than the existing design basis accidents considering additional failures. Design extension conditions can be derived based on engineering judgments, deterministic analysis or probabilistic analysis of the nuclear power plants. They are used to secure practical response capabilities to prevent or mitigate accidents. They may also require the deployment of additional safety equipment for existing nuclear power plants currently in operation. Though the general requirements of design extension conditions are described under the IAEA standards, no specific guidelines have been presented as required for their actual application to the nuclear power plant design. Furthermore, there is great variation between countries in implementing the requirements of design extension conditions. Therefore, for the actual application, considerable effort should be made among relevant organizations to establish detailed requirements of the design extension conditions. Such activities could constitute a part of the efforts of the nuclear community to meet the general public's expectations concerning the safety of nuclear power plants. The introduction of design extension conditions is expected to be a means of systematically enhancing the safety of nuclear power plants. Yet, there exists great differences in terms of the scope of analysis and the acceptance criteria, as no uniform practices have yet been established in applying the specific requirements for design extension conditions. A careful review is required in terms of the technical basis for setting the requirements, including those pertaining to the scope of analysis and the acceptance criteria. The introduction of these new requirements to Korean nuclear power plants may cause unexpected problems. Therefore, it is desirable for the regulatory agency to systematically assess the impact of design extension conditions and to discuss the arising issues with the stake holder

Philosophy for seismic design of nuclear power plants

International Nuclear Information System (INIS)

Teramae, Tetsuo

1981-01-01

In Japan, earthquakes occur frequently, therefore the basic philosophy in the aseismatic design of nuclear facilities is to design so as not to cause the accident which gives to the public in the surroundings and the employes radiation injuries in the case of large earthquakes. The ''Guideline for the aseismatic design techniques for nuclear power stations'' was drawn up in 1970 as the result of studies by related government offices and organizations. The guideline for determining the earthquakes used for design was published later, and the allowable stress for equipments and pipings has been adopted in accordance with ASME Code, Section 3. The buildings and structures, equipments and pipings in nuclear facilities are classified into three classes according to their importance in aseismatic design. The power of design earthquakes is determined corresponding to the degree of importance. The determination of the standard earthquake waves is explained. The proprieth of aseismatic design is evaluated on the basis of the basic concept of the combination of loads and the allowable limit. The static analysis in accordance with the Building Standards Act is applied to the B and C classes, while the dynamic analysis is required for the A class. The aseismatic analysis of buildings and structures, equipments and pipings is outlined. Many problems to be solved still remain though the concept of aseismatic design has been clarified. (Kako, I.)

ITER technical meeting on nuclear analysis

International Nuclear Information System (INIS)

Khripunov, V.

2000-01-01

The ITER technical meeting on nuclear analysis was organized on 24-25 February 2000 at the ITER Joint Work Site in Garching. It was clear from the meeting that continuous nuclear analysis is a fundamental part of the design process

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

Nuclear design and analysis for PWR

International Nuclear Information System (INIS)

Lee, C.K.; Kim, J.S.; Lee, S.K.; Moon, K.S.; Chun, B.J.; Chang, J.W.

1981-01-01

The list of the developed code family in this Institute has been increased after having developed two linkage codes, namely, SHUFFLE/KIDD and LEOTOKID. The former can be harnessed to supply input burnup history data to KIDD being based on the reloading patterns at the beginning of each cycle using the concentration file of KIDD, whereas the latter is able to supply the group constants to KIDD directly from the calculated results of LEOKARD by means of tapes or disks. DOT and KENO are selected specifically for benchmarking the design methods and procedures of the nuclear design codes. On the other hand, AMPX Modular code systems have been adopted for the generation of fine-or broad-group cross-sections for these benchmark codes. Language conversion and modifications of AMPX Module are taking place at the present time

Structural design and dynamic analysis of underground nuclear reactor containments

International Nuclear Information System (INIS)

Kierans, T.W.; Reddy, D.V.; Heale, D.G.

1975-01-01

Present actual experience in the structural design of undeground containments is limited to only four rather small reactors all located in Europe. Thus proposals for future underground reactors depend on the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. An application of such considerations in a recent input for the Underground Containment sub-section of the Seismic Task Group Report to the ASCE Committee for Nuclear Structures and Materials is presented as follows: underground concept considerations, siting criteria and structural selection, structural types, analytical and semi-analytical approaches, design and other miscellaneous considerations

Safety design of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

1984-01-01

Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

Joint application of AI techniques, PRA and disturbance analysis methodology to problems in the maintenance and design of nuclear power plants

International Nuclear Information System (INIS)

Okrent, D.

1989-01-01

This final report summarizes the accomplishments of a two year research project entitled ''Joint Application of Artificial Intelligence Techniques, Probabilistic Risk Analysis, and Disturbance Analysis Methodology to Problems in the Maintenance and Design of Nuclear Power Plants. The objective of this project is to develop and apply appropriate combinations of techniques from artificial intelligence, (AI), reliability and risk analysis and disturbance analysis to well-defined programmatic problems of nuclear power plants. Reactor operations issues were added to those of design and maintenance as the project progressed

Joint application of AI techniques, PRA and disturbance analysis methodology to problems in the maintenance and design of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Okrent, D.

1989-03-01

This final report summarizes the accomplishments of a two year research project entitled Joint Application of Artificial Intelligence Techniques, Probabilistic Risk Analysis, and Disturbance Analysis Methodology to Problems in the Maintenance and Design of Nuclear Power Plants. The objective of this project is to develop and apply appropriate combinations of techniques from artificial intelligence, (AI), reliability and risk analysis and disturbance analysis to well-defined programmatic problems of nuclear power plants. Reactor operations issues were added to those of design and maintenance as the project progressed.

Existing methodologies in the design and analysis of offshore floating nuclear power plants

International Nuclear Information System (INIS)

Thangam Babu, P.V.; Reddy, D.V.

1977-01-01

The paper presents a comprehensive state-of-the-art on the design and analysis of Floating Nuclear Power Plants (FNPs). The recent accelerated growth of the offshore oil industry has considerably increased the confidence in the offshore FNP concept, in view of the vast potential for the transposition of available technology. The main advantages of FNPs are: (1) unlimited supply of the cooling water, (2) isolation of thermal, noise and radioactive pollution, (3) elimination of the need for large areas of unoccupied lands usually required for safety precaution, and (4) financial savings by using standardised design and production line approach. The topics covered in this paper are: Offshore Concept Evaluations; Siting Considerations; Design Considerations; Analysis; and Miscellaneous Considerations. (Fatigue and crack propagation, Model Studies and experimental investigation, Seismic instrumentation, Noise and vibration level considerations, Safety). A detailed bibliography is presented to indicate the immediate need for further research in the areas of dynamic analysis using improved mathematical modelling techniques incorporating water-structure interaction, and nonlinear effects of the supporting medium and the mooring system; safety analysis of FNPs to accident

Nuclear power plant system environmental design and decision methodology

International Nuclear Information System (INIS)

Zendehrouh, Z.; Shinozuka, M.; Schauer, F.P.

1975-01-01

The methodology described is concerned with a system reliability analysis by which the correlation among the level of design for the environmental and natural phenomena (earthquake, flood, tornado, etc.), reasonable practical measure of safety (such as conventional safety factor), and damage (radioactivity release) probability are established. In fact, the methodology indicates how the risk of environmental and natural hazard is combined with a specific design in order to evaluate damage probability associated with the design. This leads to the optimum design decision when combined further with the cost considerations involving the radioactivity release. This fundamental approach is essential in the design of nuclear plant structures, because, unlike the convential structures, the architectural considerations and structural analysis requirements alone cannot, by themselves, result in a balanced design in the framework of social requirements. The proposed methodology incorporates the different methods of environmental load determinations with their respective probabilistic formulations as well as detailed and advanced multi-discipline (structural, mechanical, soil, nuclear physics, biology, etc.) theoretical and empirical analysis including the effect of probabilistic nature of design variables, to establish a sound and reasonable design decision model for nuclear power plants. The information required for the analysis is also described and the areas for which further research is desirable are pointed out. Furthermore, the proposed methodology can very well be utilized to determine the requirements of standardized plants to facilitate the speed of their design and review process

Nuclear-Powered GPS Spacecraft Design Study

Energy Technology Data Exchange (ETDEWEB)

Raab, Bernard

1977-05-01

This is the final report of a study to investigate the potential benefits of a nuclear (radioisotope) - powered satellite for advanced phases of the Global Positioning System (GPS) program. The critical parameters were: power to user; mean mission duration; orbital predictability; thermal control of on-board frequency standards; and vulnerability. The reference design approach is described, and input data are given for two power systems that are under development: an organic Rankine system and a Brayton cycle system. Reference design details are provided and structural design and analysis are discussed, as well as thermal design and analysis. A higher altitude version is also considered.

Research on design of dialogue session concerning nuclear technology

International Nuclear Information System (INIS)

Karikawa, Daisuke; Takahashi, Makoto; Yagi, Ekou; Toba, Tae; Kitamura, Masaharu

2010-01-01

Toward resolving nuclear conflict, deep dialogue sessions among stakeholders having different opinions concerning nuclear technology are strongly required. In order to realize a fair and constructive dialogue session, it is necessary to pay attention to the design of the dialogue session, e.g., participant selection, agenda setting, facilitation, and rules of dialogue. The basic requirements for a dialogue session have been proposed in this study based on theoretical and empirical analyses of previous dialogue sessions on nuclear issues. The theoretical analysis has been performed based on various aspects concerning defects in nuclear communication mentioned in science and technology studies. The empirical analysis has been performed by analyzing participants' responses and by systematizing practical findings of previous dialogues. The proposed requirements for a dialogue session were utilized for the design and operation of a preliminary attempt of a dialogue session named 'Open Forum for Nuclear Communication.' Through the analysis of data, such as minutes and recorded conversations collected after the session, it has been confirmed that the prespecified requirements have been satisfied. In addition, the participants' responses have shown a high acceptance for an open forum. According to these results, the basic validity and effectiveness of the proposed requirements in the design of a dialogue session have been successfully demonstrated. (author)

Open-Source Integrated Design-Analysis Environment For Nuclear Energy Advanced Modeling & Simulation Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

O' Leary, Patrick [Kitware, Inc., Clifton Park, NY (United States)

2017-01-30

The framework created through the Open-Source Integrated Design-Analysis Environment (IDAE) for Nuclear Energy Advanced Modeling & Simulation grant has simplify and democratize advanced modeling and simulation in the nuclear energy industry that works on a range of nuclear engineering applications. It leverages millions of investment dollars from the Department of Energy's Office of Nuclear Energy for modeling and simulation of light water reactors and the Office of Nuclear Energy's research and development. The IDEA framework enhanced Kitware’s Computational Model Builder (CMB) while leveraging existing open-source toolkits and creating a graphical end-to-end umbrella guiding end-users and developers through the nuclear energy advanced modeling and simulation lifecycle. In addition, the work deliver strategic advancements in meshing and visualization for ensembles.

Nuclear component design ontology building based on ASME codes

International Nuclear Information System (INIS)

Bao Shiyi; Zhou Yu; He Shuyan

2005-01-01

The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)

Evaluation of masonry wall design at nuclear power plants

International Nuclear Information System (INIS)

Con, V.N.; Subramonian, N.; Chokshi, N.

1983-01-01

The structural integrity of safety-related masonry walls in operating nuclear power plants may not be maintained when subjected to certain loads and load combinations. The paper presents some findings based upon the review of the design and analysis procedures used by the licensees in the reevaluation of safety-related masonry walls. The design criteria developed by the Structural Engineering Branch (SEB) of the United States Nuclear Regulatory Commission (NRC) along with other standard codes such as the Uniform Building Code, ACI 531-79, ATC 3-06, and NCMA were used as guidance in evaluating the design criteria developed by the licensees. The paper deals with the following subject areas: loads and load combinations, allowable stresses, analytical procedures, and modification methods. The paper concludes that, in general, the masonry walls in nuclear power plants comply with the working stress design requirements. In some cases, certain nonlinear analysis methods were used. The applicability of these methods is discussed. (orig.)

Seismic analysis of nuclear power plants

International Nuclear Information System (INIS)

Halbritter, A.L.

1984-01-01

Nuclear Power Plants require exceptional safety guarantees which are reflected in a rigorous control of the employed materials, advanced construction technology, sophisticated methods of analysis and consideration of non conventional load cases such as the earthquake loading. In this paper, the current procedures used in the seismic analysis of Nuclear Power Plants are presented. The seismic analysis of the structures has two objectives: the determination of forces in the structure in order to design it against earthquakes and the generation of floor response spectra to be used in the design of mechanical and electrical components and piping systems. (Author) [pt

Technical guidelines for aseismic design of nuclear power plants

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.

1994-06-01

This document is a translation, in its entirety, of the Japan Electric Association (JEA) publication entitled open-quotes Technical Guidelines for Aseismic Design of Nuclear Power Plants - JEAG 4601-1987.close quotes This guideline describes in detail the aseismic design techniques used in Japan for nuclear power plants. It contains chapters dealing with: (a)the selection of earthquake ground motions for a site, (b) the investigation of foundation and bedrock conditions, (c) the evaluation of ground stability and the effects of ground movement on buried piping and structures, (d) the analysis and design of structures, and (e) the analysis and design of equipment and distribution systems (piping, electrical raceways, instrumentation, tubing and HVAC duct). The guideline also includes appendices which summarize data, information and references related to aseismic design technology

Analysis of color environment in nuclear power plants

International Nuclear Information System (INIS)

Natori, Kazuyuki; Akagi, Ichiro; Souma, Ichiro; Hiraki, Tadao; Sakurai, Yukihiro.

1996-01-01

This article reports the results of color and psychological analysis of the outlook of nuclear power plants and the visual environments inside of the plants. Study one was the color measurements of the outlook of nuclear plants and the visual environment inside of the plants. Study two was a survey of the impressions on the visual environments of nuclear plants obtained from observers and interviews of the workers. Through these analysis, we have identified the present state of, and the problems of the color environments of the nuclear plants. In the next step, we have designed the color environments of inside and outside of the nuclear plants which we would recommend (inside designs were about fuel handling room, operation floor of turbine building, observers' pathways, central control room, rest room for the operators). Study three was the survey about impressions on our design inside and outside of the nuclear plants. Nuclear plant observers, residents in Osaka city, residents near the nuclear plants, the operators, employees of subsidiary company and the PR center guides rated their impressions on the designs. Study four was the survey about the design of the rest room for the operators controlling the plants. From the results of four studies, we have proposed some guidelines and problems about the future planning about the visual environments of nuclear power plants. (author)

Technical conservatism in the design and analysis of a nuclear-waste repository in basalt

International Nuclear Information System (INIS)

Jones, K.A.

1982-01-01

The US Department of Energy's National Waste Terminal Storage Program has adopted a policy of technical conservatism to guide the design and analysis of geologic disposal systems for commercial high-level radioactive waste. Technical conservatism serves as the programmatic philosophy for managing uncertainty in the performance of the disposal system. The implementation of technical conservatism as applied to a nuclear waste repository in basalt is discussed. Preliminary assessments of the performance of the waste package, repository, and site subsystems are compared to key proposed regulatory criteria. The comparison shows that there are substantial safety margins in the predicted performance of the nuclear waste repository in basalt

Analysis and design of a nuclear boiler control scheme

International Nuclear Information System (INIS)

Hughes, F.M.

1982-01-01

The steam raising plant of a British designed nuclear power station for which the installed control scheme has given rise to performance and stability problems is considered. The lecture is based on studies carried out for the plant which illustrate the way in which multivariable frequency response methods can be used to analyse and identify the source of control problems and further enable alternative control schemes, having improved performance, to be designed. (author)

The design analysis of ACP-canister for nuclear waste disposal

International Nuclear Information System (INIS)

Raiko, H.

1992-05-01

The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

Advanced nuclear energy analysis technology

International Nuclear Information System (INIS)

Gauntt, Randall O.; Murata, Kenneth K.; Romero, Vicente Josce; Young, Michael Francis; Rochau, Gary Eugene

2004-01-01

A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems

Facility Safeguardability Analysis In Support of Safeguards-by-Design

Energy Technology Data Exchange (ETDEWEB)

Philip Casey Durst; Roald Wigeland; Robert Bari; Trond Bjornard; John Hockert; Michael Zentner

2010-07-01

The following report proposes the use of Facility Safeguardability Analysis (FSA) to: i) compare and evaluate nuclear safeguards measures, ii) optimize the prospective facility safeguards approach, iii) objectively and analytically evaluate nuclear facility safeguardability, and iv) evaluate and optimize barriers within the facility and process design to minimize the risk of diversion and theft of nuclear material. As proposed by the authors, Facility Safeguardability Analysis would be used by the Facility Designer and/or Project Design Team during the design and construction of the nuclear facility to evaluate and optimize the facility safeguards approach and design of the safeguards system. Through a process of “Safeguards-by-Design” (SBD), this would be done at the earliest stages of project conceptual design and would involve domestic and international nuclear regulators and authorities, including the International Atomic Energy Agency (IAEA). The benefits of the Safeguards-by-Design approach is that it would clarify at a very early stage the international and domestic safeguards requirements for the Construction Project Team, and the best design and operating practices for meeting these requirements. It would also minimize the risk to the construction project, in terms of cost overruns or delays, which might otherwise occur if the nuclear safeguards measures are not incorporated into the facility design at an early stage. Incorporating nuclear safeguards measures is straight forward for nuclear facilities of existing design, but becomes more challenging with new designs and more complex nuclear facilities. For this reason, the facility designer and Project Design Team require an analytical tool for comparing safeguards measures, options, and approaches, and for evaluating the “safeguardability” of the facility. The report explains how preliminary diversion path analysis and the Proliferation Resistance and Physical Protection (PRPP) evaluation

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Design-for-analysis or the unintended role of analysis in the design of piping systems

International Nuclear Information System (INIS)

Antaki, G.A.

1991-01-01

The paper discusses the evolution of piping design in the nuclear industry with its increasing reliance on dynamic analysis. While it is well recognized that the practice has evolved from ''design-by- rule '' to ''design-by-analysis,'' examples are provided of cases where the choice of analysis technique has determined the hardware configuration, which could be called ''design-for-analysis.'' The paper presents practical solutions to some of these cases and summarizes the important recent industry and regulatory developments which, if successful, will reverse the trend towards ''design-for-analysis.'' 14 refs

Nuclear fuel cycle system analysis

International Nuclear Information System (INIS)

Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

2012-04-01

The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

Joint application of AI techniques, PRA and disturbance analysis methodology to problems in the maintenance and design of nuclear power plants. Final report

Energy Technology Data Exchange (ETDEWEB)

Okrent, D.

1989-03-01

This final report summarizes the accomplishments of a two year research project entitled ``Joint Application of Artificial Intelligence Techniques, Probabilistic Risk Analysis, and Disturbance Analysis Methodology to Problems in the Maintenance and Design of Nuclear Power Plants. The objective of this project is to develop and apply appropriate combinations of techniques from artificial intelligence, (AI), reliability and risk analysis and disturbance analysis to well-defined programmatic problems of nuclear power plants. Reactor operations issues were added to those of design and maintenance as the project progressed.

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself; be caused by human action; be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite); conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant; or non-radiological effects arising

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself. Be caused by human action. Be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite). Conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant. Or non-radiological effects arising

Nuclear Futures Analysis and Scenario Building

International Nuclear Information System (INIS)

Arthur, E.D.; Beller, D.; Canavan, G.H.; Krakowski, R.A.; Peterson, P.; Wagner, R.L.

1999-01-01

This LDRD project created and used advanced analysis capabilities to postulate scenarios and identify issues, externalities, and technologies associated with future ''things nuclear''. ''Things nuclear'' include areas pertaining to nuclear weapons, nuclear materials, and nuclear energy, examined in the context of future domestic and international environments. Analysis tools development included adaptation and expansion of energy, environmental, and economics (E3) models to incorporate a robust description of the nuclear fuel cycle (both current and future technology pathways), creation of a beginning proliferation risk model (coupled to the (E3) model), and extension of traditional first strike stability models to conditions expected to exist in the future (smaller force sizes, multipolar engagement environments, inclusion of actual and latent nuclear weapons (capability)). Accomplishments include scenario development for regional and global nuclear energy, the creation of a beginning nuclear architecture designed to improve the proliferation resistance and environmental performance of the nuclear fuel cycle, and numerous results for future nuclear weapons scenarios

Study of seismic design bases and site conditions for nuclear power plants

International Nuclear Information System (INIS)

1980-04-01

This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches

Study of seismic design bases and site conditions for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

1980-04-01

This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

Applicability of trends in nuclear safety analysis to space nuclear power systems

International Nuclear Information System (INIS)

Bari, R.A.

1992-01-01

A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication

The basic discussion on nuclear power safety improvement based on nuclear equipment design

International Nuclear Information System (INIS)

Zhao Feiyun; Yao Yangui; Yu Hao; He Yinbiao; Gao Lei; Yao Weida

2013-01-01

The safety of strengthening nuclear power design was described based on nuclear equipment design after Fukushima nuclear accident. From these aspects, such as advanced standard system, advanced design method, suitable test means, consideration of beyond design basis event, and nuclear safety culture construction, the importance of nuclear safety improvement was emphatically presented. The enlightenment was given to nuclear power designer. (authors)

Design of a nuclear-waste package for emplacement in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Rothman, A.J.; Gregg, D.W.; Hockman, J.N.; Revelli, M.A.; Russell, E.W.; Schornhorst, J.R.

1983-01-01

Design, modeling, and testing activities are under way at LLNL in the development of high level nuclear waste package designs. We discuss the geological characteristics affecting design, the 10CFR60 design requirements, conceptual designs, metals for containment barriers, economic analysis, thermal modeling, and performance modeling

NUCLEAR SAFETY DESIGN BASES FOR LICENSE APPLICATION

International Nuclear Information System (INIS)

Garrett, R.J.

2005-01-01

The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

Design, maintenance and lifetime of nuclear components

International Nuclear Information System (INIS)

Noel, R.L.; Eisenhut, D.G.; Carey, J.J.; Reynes, L.J.

1989-01-01

Division D of SMiRT deals with experience feedback relating to the in-service behavior of nuclear components, the design and construction of this equipment, its maintenance and the evaluation and management of its lifetime. The nuclear industry now having reached maturity, with more than 300 units in service worldwide, these problems are now of predominant importance to the activity of the industry and in its development programs. This applies particularly to the problems relating to the lifetime of nuclear plants, problems which are rightly of such concern both to the utilities, in view of the enormous investments involved, and also to the safety authorities. These contributions have been reviewed for the purpose of analyzing the essential points. This analysis highlights the considerable advances achieved during the recent decades in design and maintenance methods and practices. It also identifies the areas in which progress still remains to be made

Topics on Japanese aseismic design for nuclear installations

International Nuclear Information System (INIS)

Nakamura, Masahiko

2002-01-01

Major items of of Japanese anti-seismic design for nuclear installations involve three topics: earthquakes and ground motion; seismic design and safety evaluation. The first topic deal with: improvement of geological survey technology, evaluation of ground motion from active faults, and characterisation of earthquake from individual faults. Seismic design involves: evaluation of design ground motion (spectra), classification of structures, systems and components (SSCs) based on the seismic importance, and seismic design criteria and critical loads. Safety evaluation of seismic PSA is dependent on the consistency of the the two previous items. Seismic hazard evaluation methodology, database and examples of analysis are described. Analysis method using fault model is included

Design and implementation of the system of importing data of nuclear analysis experiment based on UML technology

International Nuclear Information System (INIS)

Peng Yanguang; Yu Chuansong; Ma Mei; Sun Gongxing

2004-01-01

There are some problems that Excel data importing to database for our scientific data processing, we introduce a method how to use UML and XML technology design and implementation our system. Because XML technology have flexible data exchange mechanism, so it is appropriate to do it, and it is easy for use that the system of importing data of nuclear analysis experiment has a web browser GUI. Our paper emphasis on the software design method and UML design cases. (authors)

Aseismatic design and safety of nuclear power generation facilities. On aseismatic capability of commercial nuclear power stations

International Nuclear Information System (INIS)

Kato, Muneaki

1995-01-01

In view of the great Hanshin earthquake, the aseismatic safety of the important facilities in nuclear power stations is ensured by the location direct on base rocks, the design with the earthquake force at least three times as large as that in the building standard, and the consideration of the earthquakes due to active faults as design earthquake. The basic policy of the aseismatic design of nuclear power stations is described. The determination of the earthquake motions due to strongest earthquake and utmost limit earthquake for design, the survey of the geological features and ground of the sites and so on are explained. In the aseismatic design of buildings and structures, structural planning, the modeling for the aseismatic analysis of buildings, the analysis of time historical response and so on are carried out. In the aseismatic design of equipment and piping systems, the planning of aseismatic support structures, the aseismatic design and the analysis of time historical response, the spectral modal analysis for other systems such as multiple material point system and so on are described. The tests and researches related to the aseismatic design are reported. (K.I.)

Investigation on prediction capability of nuclear design parameters for gap configuration in ITER through analysis of the FNS gap streaming experiment

International Nuclear Information System (INIS)

Maekawa, Fujio; Konno, Chikara; Kasugai, Yoshimi; Oyama, Yukio; Uno, Yoshitomo; Maekawa, Hiroshi; Ikeda, Yujiro

2000-01-01

As an R and D Task of shielding neutronics experiment under the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), streaming experiments with simulating a gap configuration formed by two neighboring blanket modules of ITER were carried out at the FNS (Fusion Neutron Source) facility. In this work, prediction capability of various nuclear design parameters was investigated through analysis of the experiments. The Monte Carlo transport calculation code MCNP-4A and the FENDL/E-1.0 and JENDL Fusion File cross section data libraries were used for the analysis with detailed modeling of the experimental conditions. As a result, all the measured quantities were reproduced within about ±30% by the calculations. It was concluded that these calculation tools were capable of predicting nuclear design parameters, such as helium production rates at connection legs of blanket modules to the back plate and nuclear responses in toroidal field coils, with uncertainty of ±30% for the geometry where gap-streaming effect was significant. (author)

Rethinking Sensitivity Analysis of Nuclear Simulations with Topology

Energy Technology Data Exchange (ETDEWEB)

Dan Maljovec; Bei Wang; Paul Rosen; Andrea Alfonsi; Giovanni Pastore; Cristian Rabiti; Valerio Pascucci

2016-01-01

In nuclear engineering, understanding the safety margins of the nuclear reactor via simulations is arguably of paramount importance in predicting and preventing nuclear accidents. It is therefore crucial to perform sensitivity analysis to understand how changes in the model inputs affect the outputs. Modern nuclear simulation tools rely on numerical representations of the sensitivity information -- inherently lacking in visual encodings -- offering limited effectiveness in communicating and exploring the generated data. In this paper, we design a framework for sensitivity analysis and visualization of multidimensional nuclear simulation data using partition-based, topology-inspired regression models and report on its efficacy. We rely on the established Morse-Smale regression technique, which allows us to partition the domain into monotonic regions where easily interpretable linear models can be used to assess the influence of inputs on the output variability. The underlying computation is augmented with an intuitive and interactive visual design to effectively communicate sensitivity information to the nuclear scientists. Our framework is being deployed into the multi-purpose probabilistic risk assessment and uncertainty quantification framework RAVEN (Reactor Analysis and Virtual Control Environment). We evaluate our framework using an simulation dataset studying nuclear fuel performance.

Nuclear Safety Design Base for License Application

International Nuclear Information System (INIS)

R.J. Garrett

2005-01-01

The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

The computer program system for structural design of nuclear power plants

International Nuclear Information System (INIS)

Aihara, S.; Atsumi, K.; Sasagawa, K.; Satoh, S.

1979-01-01

In recent days, the design method of the Nuclear Power Plant has become more complex than in the past. The Finite Element Method (FEM) applied for analysis of Nuclear Power Plants, especially requires more computer use. The recent computers have made remarkable progress, so that in design work manpower and time necessary for analysis have been reduced considerably. However, instead the arrangement of outputs have increased tremendously. Therefore, a computer program system was developed for performing all of the processes, from data making to output arrangement, and rebar evaluations. This report introduces the computer program system pertaining to the design flow of the Reactor Building. (orig.)

Nuclear waste package design for the Vadose zone in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Ballou, L.B.; Gregg, D.W.; Russell, E.W.

1984-02-01

This report presents an overview of the selection and analysis of conceptual waste package designs that will be used by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for disposal of high-level nuclear waste (HLW) at the proposed Yucca Mountain, Nevada Site. The design requirements that the waste packages are required to meet are listed. Concept drawings for the reference designs and one alternative package design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and Incoloy 825 have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and economic analysis supporting the selection of the conceptual waste package designs is included. Post-closure containment and release rates are not discussed in this paper. 17 references, 2 figures, 2 tables

Human factor engineering applied to nuclear power plant design

International Nuclear Information System (INIS)

Manrique, A.; Valdivia, J.C.

2007-01-01

Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. Such a plan should state: -) Activities to be performed, and -) Creation of a Human Factor Engineering team adequately qualified. The Human Factor Engineering team is an integral part of the design team and is strongly linked to the engineering organizations but simultaneously has independence to act and is free to evaluate designs and propose changes in order to enhance human behavior. TECNATOM S.A. (a Spanish company) has been a part of the Design and Human Factor Engineering Team and has collaborated in the design of an advanced Nuclear Power Plant, developing methodologies and further implementing those methodologies in the design of the plant systems through the development of the plant systems operational analysis and of the man-machine interface design. The methodologies developed are made up of the following plans: -) Human Factor Engineering implementation in the Man-Machine Interface design; -) Plant System Functional Requirement Analysis; -) Allocation of Functions to man/machine; -) Task Analysis; -) Human-System Interface design; -) Control Room Verification and -) Validation

Computer aided design for the nuclear industry

International Nuclear Information System (INIS)

Basson, Keith

1986-01-01

The paper concerns the new computer aided design (CAD) centre for the United Kingdom nuclear industry, and its applications. A description of the CAD system is given, including the current projects at the CAD centre. Typical applications of the 3D CAD plant based models, stress analysis studies, and the extraction of data from CAD drawings to produce associated documentation, are all described. Future developments using computer aided design systems are also considered. (U.K.)

An estimate and evaluation of design error effects on nuclear power plant design adequacy

International Nuclear Information System (INIS)

Stevenson, J.D.

1984-01-01

An area of considerable concern in evaluating Design Control Quality Assurance procedures applied to design and analysis of nuclear power plant is the level of design error expected or encountered. There is very little published data 1 on the level of error typically found in nuclear power plant design calculations and even less on the impact such errors would be expected to have on overall design adequacy of the plant. This paper is concerned with design error associated with civil and mechanical structural design and analysis found in calculations which form part of the Design or Stress reports. These reports are meant to document the design basis and adequacy of the plant. The estimates contained in this paper are based on the personal experiences of the author. In Table 1 is a partial listing of the design docummentation review performed by the author on which the observations contained in this paper are based. In the preparation of any design calculations, it is a utopian dream to presume such calculations can be made error free. The intent of this paper is to define error levels which might be expected in a competent engineering organizations employing currently technically qualified engineers and accepted methods of Design Control. In addition, the effects of these errors on the probability of failure to meet applicable design code requirements also are estimated

Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

2015-05-15

NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

PA activity by using nuclear power plant safety demonstration and analysis

International Nuclear Information System (INIS)

Tsuchiya, Mitsuo; Kamimae, Rie

1999-01-01

INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

Status of core nuclear design technology for future fuel

International Nuclear Information System (INIS)

Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

1997-01-01

The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

Designing the owner's nuclear project management organization

International Nuclear Information System (INIS)

Cooke, T.C.; Peck, B.H.

1976-01-01

Few decisions are more important to an electric utility company than the one to build a nuclear generating facility. This decision will require continuous management attention to the nuclear project for periods of up to ten years on the part of the utility. Effective management of such a large, complex project requires an owner's organization skilled in such areas as engineering, heavy construction, procurement, and project control. The paper describes a method for designing the owner's nuclear organization. Factors considered include the identification of milestone events and phases of the project and identification of key organizational groups and their degree of involvement. A series of important decision milestones is also identified for structuring the organization. A step-by-step analysis involving a set of evaluation criteria results in a recommended organization that can be staffed by the owner according to the degree of involvement desired. This technique of analysis could also be performed using different evaluation criteria resulting in other options for the owner's organization

Analysis of near optimum design for small and medium size nuclear power plants

International Nuclear Information System (INIS)

Ahmed, A.A.

1977-01-01

Market surveys in recent years have shown that a significant market would exist among the developing nations of the world for nuclear power plants that would be classified as small to medium sized, provided that these small plants could produce electricity at a unit price comparable to that of equivalent sized fossil fired plants. Nuclear plants in the range of 100 MWe to 500 MWe would fit more effectively into the relatively smaller grids of most developing nations than would the 900 MWe to 1300 MWe units now being constructed in the large industrial nations. Worldwide re-evaluation of the worth of fossil fuels has prompted a re-examination of the competitive position of small to medium sized nuclear generating units compared to comparable fossil fired units, especially in the context of units specifically optimized for the size range of interest, rather than of designs that are simply scaled down versions of the currently available larger units. Since the absolute cost of electricity is more sensitive to external factors such as cost of money, national inflation rate and time required for licensing and construction than to details of design or perhaps even to choice of fuels, and since the cost of electricity generated in small to medium sized fossil fired units is periodically compared to that of scaled down versions of conventional large nuclear units, the point of view taken here is one of comparing the relative generating costs of smaller nuclear units of optimum design with the corresponding costs of scaled down versions of current large nuclear generating units

OLP embedment design method research for AP1000 nuclear plant

International Nuclear Information System (INIS)

Li Cheng; Li Shaoping; Liu Jianwei

2013-01-01

Background: One of the most advanced nuclear power technology, the first AP1000 reactor is under construction in China. Modularization is one of the main characteristics for AP1000 nuclear plant building. Module wall with steel face plate is used instead of reinforced concrete structure wall. A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes, equipment, operation platforms and any other component attached to the module wall. Therefore, the design of embedment is very important in AP1000 structural design. Purpose: A finite element analysis method and tool for embedment design is needed for convenience. Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design. Results: A Microsoft Excel template for embedment design is developed. Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice. (authors)

Status of the ITER tokamak nuclear shielding and radiological protection design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

2016-11-01

Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

International Nuclear Information System (INIS)

Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

2016-01-01

Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

Energy Technology Data Exchange (ETDEWEB)

Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

2016-12-15

Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

International Nuclear Information System (INIS)

Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

2011-01-01

Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

Designing new nuclear chemical processing plants for safeguards accountability

International Nuclear Information System (INIS)

Sprouse, K.M.

1987-01-01

New nuclear chemical processing plants will be required to develop material accountability control limits from measurement error propagation analysis rather than historical inventory difference data as performed in the past. In order for measurement error propagation methods to be viable alternatives, process designers must ensure that two nondimensional accountability parameters are maintained below 0.1. These parameters are ratios between the material holdup increase and the variance in inventory difference measurement uncertainty. Measurement uncertainty data for use in error propagation analysis is generally available in the open literature or readily derived from instrument calibration data. However, nuclear material holdup data has not been adequately developed for use in the material accountability design process. Long duration development testing on isolated unit operations is required to generate this necessary information

Design basis threat analysis and implementation of the physical protection system at Nuclear Facility of BATAN Yogyakarta

International Nuclear Information System (INIS)

Syarip

2005-01-01

An analysis to determine the design basis threat (DBT) and its follow-up through the implementation of physical protection system at the nuclear facility of BATAN Yogyakarta has been done. Methodology used for the analysis is based on the IAEA guidance for the development and maintenance of a DBT. Based on the analysis results, it can be concluded that the threat motivation is influenced by political situation (related to the government policy), criminal, sabotage and theft. The characteristics of threats are: not so well organized, terror, theft of materials information, involving insider (collusion), and intimidation to workers. Potential threat could from guests/students who take a practical job or laboratory exercise. Therefore, it is necessary to be anticipated the possibility and its impact of turmoil/demonstrators such as destruction of: lighting, road, fence, sabotage on the electric and communication lines, surrounding the Yogyakarta nuclear facility

Preliminary design study for a carbide LEU-nuclear thermal rocket

International Nuclear Information System (INIS)

Venneri, P.F.; Kim, Y.

2014-01-01

Nuclear space propulsion is a requirement for the successful exploration of the solar system. It offers the possibility of having both a high specific impulse and a relatively high thrust, allowing rapid transit times with a minimum usage of fuel. This paper proposes a nuclear thermal rocket design based on heritage NERVA rockets that makes use of Low Enriched Uranium (LEU) fuel. The Carbide LEU Nuclear Thermal Rocket (C-LEU-NTR) is designed to fulfill the rocket requirements as set forth in the NASA 2009 Mars Mission Design Reference Architecture 5.0, that is provide 25,000 lbf of thrust, operate at full power condition for at least two hours, and have a specific impulse close to 900 s. The neutronics analysis was done using MCNP5 with the ENDF/B-VII.1 neutron library. The thermal hydraulic calculations and size optimization were completed with a finite difference code being developed at the Center for Space Nuclear Research. (authors)

Design of equipment management information system for nuclear power plant

International Nuclear Information System (INIS)

Wang Chengyuan

1996-01-01

The author describes the ideas and practical method for need analysis, system function dividing, code design, program design and network disposition of equipment purchase management system of nuclear power plant during building, from the view of engineering investment control, schedule control and quality control

Holistic safety analysis for advanced nuclear power plants

International Nuclear Information System (INIS)

Alvarenga, M.A.B.; Guimaraes, A.C.F.

1992-01-01

This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

Fundamental principles for a nuclear design and structural analysis code for HTR components operating at temperatures above 8000C

International Nuclear Information System (INIS)

Nickel, H.; Schubert, F.

1985-01-01

With reference to the special characteristics of an HTR plant for the supply of nuclear process heat, the investigation of the fundamental principles to form the basis for a high temperature nuclear structural design code has been described. As examples, preliminary design values are proposed for the creep rupture and fatigue behaviour. The linear damage accumulation rule is for practical reasons proposed for the determination of service life, and the difficulties in using this rule are discussed. Finally, using the data obtained in structural analysis, the main areas of investigation which will lead to improvements in the utilization of the materials are discussed. Based on the current information, the working group ''Design Code'' believes that a service life of 70000 h for the heat-exchanging components operating at above 800 0 C can be. (orig.)

Nuclear waste repository design and construction

International Nuclear Information System (INIS)

Bohlke, B.M.; Monsees, J.E.

1987-01-01

Extensive underground excavation will be required for construction of a mined geologic repository for nuclear waste. Hundreds of thousands of feet of drift will be required based on the conceptual layout design for each candidate nuclear waste repository. Comparison of boring and blasting excavation methods are discussed, as are special design and construction requirements (e.g., quality assurance procedures and performance assessment) for the nuclear waste repository. Comparisons are made between boring and blasting construction methods for the repository designs proposed for salt, volcanic tuff, and basalt

Nuclear design characteristics of SMART

Energy Technology Data Exchange (ETDEWEB)

Lee, Chung Chan; Park, Sang Yoon; Lee, Ki Bog; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

Nuclear design bases for System-Integrated Modular Advanced ReacTor(SMART) core are presented. Based on the proposed design bases, a SMART core loading pattern is constructed and its nuclear characteristics are studied. The proposed core loading pattern satisfies 3-year cycle length and soluble boron-free operation requirements at any time during the cycle. 10 refs., 2 figs., 1 tab. (Author)

Nuclear design characteristics of SMART

Energy Technology Data Exchange (ETDEWEB)

Lee, Chung Chan; Park, Sang Yoon; Lee, Ki Bog; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Nuclear design bases for System-Integrated Modular Advanced ReacTor(SMART) core are presented. Based on the proposed design bases, a SMART core loading pattern is constructed and its nuclear characteristics are studied. The proposed core loading pattern satisfies 3-year cycle length and soluble boron-free operation requirements at any time during the cycle. 10 refs., 2 figs., 1 tab. (Author)

NPP-Nuclear Island Design. From conceptual design to Project execution

International Nuclear Information System (INIS)

Lanchet, Dominique

2014-01-01

The second day opened with the lecture of Dominique Lanchet, Design Senior Vice President at AREVA Engineering and Project. Dominique Lanchet gave us an overview of the steps of a Nuclear Island Design creation from the conceptual design to the project execution, giving the examples of the EPR and ATMEA1 TM nuclear reactors

History of Nuclear Weapons Design and Production

Science.gov (United States)

Oelrich, Ivan

2007-04-01

The nuclear build-up of the United States and the Soviet Union during the Cold War is often portrayed as an arms race. Some part was indeed a bilateral competition, but much was the result of automatic application of technical advances as they became available, without careful consideration of strategic implications. Thus, the history of nuclear weapon design is partly designers responding to stated military needs and partly the world responding to constant innovations in nuclear capability. Today, plans for a new nuclear warhead are motivated primarily by the desire to maintain a nuclear design and production capability for the foreseeable future.

Materials analysis with a nuclear microprobe

International Nuclear Information System (INIS)

Maggiore, C.J.

1980-01-01

The ability to produce focused beams of a few MeV light ions from Van de Graaff accelerators has resulted in the development of nuclear microprobes. Rutherford backscattering, nuclear reactions, and particle-induced x-ray emission are used to provide spatially resolved information from the near surface region of materials. Rutherford backscattering provides nondestructive depth and mass resolution. Nuclear reactions are sensitive to light elements (Z < 15). Particle-induced x-ray analysis is similar to electron microprobe analysis, but 2 orders of magnitude more sensitive. The focused beams are usually produced with specially designed multiplets of magnetic quadrupoles. The LASL microprobe uses a superconducting solenoid as a final lens. The data are acquired by a computer interfaced to the experiment with CAMAC. The characteristics of the information acquired with a nuclear microprobe are discussed; the means of producing the beams of nuclear particles are described; and the limitations and applications of such systems are given

MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

International Nuclear Information System (INIS)

Sugino, K.; Iwai, T.

2006-01-01

MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

Concept Design and Risk Assessment of Nuclear Propulsion Ship

International Nuclear Information System (INIS)

Gil, Youngmi; Yoo, Seongjin; Kim, Yeontae; Oh, June; Byun, Yoonchul; Woo, Ilguk; Kim, Jiho; Choi, Suhn

2014-01-01

The nuclear propulsion ships (hereinafter referred to as 'nuclear ships') have been considered as an eco-friendly ship. There have historically been warship and submarine with the source of nuclear power. The use of nuclear ships has been recently extending to the icebreaker, the deep-water exploration ship, and the floating nuclear power plant. Prior to developing the new ship, we evaluated the economics of various types of ships and concluded that the container ship could be appropriate for the nuclear propulsion. In order to verify its safety, we performed the ship calculation based on the optimal arrangement of the nuclear reactor. Finally, we verified its safety by the HAZID. In the former research, we confirmed the applicability of the nuclear propulsion system for the large container ship. In this study, we verified the safety of the nuclear ships according to the HAZID analysis. We expect that this research will lead to safe design of the nuclear ships

Concept Design and Risk Assessment of Nuclear Propulsion Ship

Energy Technology Data Exchange (ETDEWEB)

Gil, Youngmi; Yoo, Seongjin; Kim, Yeontae; Oh, June; Byun, Yoonchul; Woo, Ilguk [Daewoo Shipbuilding and Marine Engineering Co. Ltd., Seoul (Korea, Republic of); Kim, Jiho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The nuclear propulsion ships (hereinafter referred to as 'nuclear ships') have been considered as an eco-friendly ship. There have historically been warship and submarine with the source of nuclear power. The use of nuclear ships has been recently extending to the icebreaker, the deep-water exploration ship, and the floating nuclear power plant. Prior to developing the new ship, we evaluated the economics of various types of ships and concluded that the container ship could be appropriate for the nuclear propulsion. In order to verify its safety, we performed the ship calculation based on the optimal arrangement of the nuclear reactor. Finally, we verified its safety by the HAZID. In the former research, we confirmed the applicability of the nuclear propulsion system for the large container ship. In this study, we verified the safety of the nuclear ships according to the HAZID analysis. We expect that this research will lead to safe design of the nuclear ships.

Time-series-analysis techniques applied to nuclear-material accounting

International Nuclear Information System (INIS)

Pike, D.H.; Morrison, G.W.; Downing, D.J.

1982-05-01

This document is designed to introduce the reader to the applications of Time Series Analysis techniques to Nuclear Material Accountability data. Time series analysis techniques are designed to extract information from a collection of random variables ordered by time by seeking to identify any trends, patterns, or other structure in the series. Since nuclear material accountability data is a time series, one can extract more information using time series analysis techniques than by using other statistical techniques. Specifically, the objective of this document is to examine the applicability of time series analysis techniques to enhance loss detection of special nuclear materials. An introductory section examines the current industry approach which utilizes inventory differences. The error structure of inventory differences is presented. Time series analysis techniques discussed include the Shewhart Control Chart, the Cumulative Summation of Inventory Differences Statistics (CUSUM) and the Kalman Filter and Linear Smoother

A new approach to nuclear reactor design optimization using genetic algorithms and regression analysis

International Nuclear Information System (INIS)

Kumar, Akansha; Tsvetkov, Pavel V.

2015-01-01

Highlights: • This paper presents a new method useful for the optimization of complex dynamic systems. • The method uses the strengths of; genetic algorithms (GA), and regression splines. • The method is applied to the design of a gas cooled fast breeder reactor design. • Tools like Java, R, and codes like MCNP, Matlab are used in this research. - Abstract: A module based optimization method using genetic algorithms (GA), and multivariate regression analysis has been developed to optimize a set of parameters in the design of a nuclear reactor. GA simulates natural evolution to perform optimization, and is widely used in recent times by the scientific community. The GA fits a population of random solutions to the optimized solution of a specific problem. In this work, we have developed a genetic algorithm to determine the values for a set of nuclear reactor parameters to design a gas cooled fast breeder reactor core including a basis thermal–hydraulics analysis, and energy transfer. Multivariate regression is implemented using regression splines (RS). Reactor designs are usually complex and a simulation needs a significantly large amount of time to execute, hence the implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using RS in conjunction with GA. Due to using RS, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a multivariate regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The reactor parameters are given by the, radius of a fuel pin cell, isotopic enrichment of the fissile material in the fuel, mass flow rate of the coolant, and temperature of the coolant at the core inlet. And, the optimization objectives for the reactor core are, high breeding of U-233 and Pu-239 in

Methods for air cleaning system design and accident analysis

International Nuclear Information System (INIS)

Gregory, W.S.; Nichols, B.D.

1987-01-01

This paper describes methods, in the form of a handbook and five computer codes, that can be used for nuclear facility air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described

Design criteria development for the structural stability of nuclear waste repository

Energy Technology Data Exchange (ETDEWEB)

Yun, C H [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Yu, T S [Daewoo Engineering Company, Sungnam (Korea, Republic of); Ko, H M [Seoul National Univ., Seoul (Korea, Republic of)

1990-11-15

The objective of the present project is to develop design criteria for the structural stability of rock cavity for the underground repository are defined, according to which detailed descriptions for design methodologies, design stages and stability analysis of the cavity are made. The proposed criteria can be used as a guide for the preparation of design codes which are to be established as the site condition and technical emplacement procedure are fixed. The present report first reviews basic safety requirements and criteria of the underground disposal of nuclear wastes for the establishment of design concepts and stability analysis of the rock cavity. Important factors for the design are also described by considering characteristics of the wastes and underground facilities. The present project has investigated technical aspects on the design of underground structures based on the currently established underground construction technologies, and presented a proposal for design criteria for the structural stability of the nuclear waste repository. The proposed criteria consist of general provisions, geological exploration, rock classification, design process and methods, supporting system, analyses and instrumentation.

Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

International Nuclear Information System (INIS)

Chen Yan; Wang Minghuang; Jiang Jieqiong

2012-01-01

A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

Developing engineering analysis capabilities at a nuclear utility

International Nuclear Information System (INIS)

Miller, J.S.

1992-01-01

When a nuclear plant is originally designed and constructed, a large staff of analytical and design personnel is used by the architectural and engineering (A/E) firm(s) and the nuclear steam supply system (NSSS) engineering firm(s) to provide the technical specifications needed for the plant to function and satisfy US Nuclear Regulatory Commission (NRC) requirements. During this design process, thousands of calculations are performed, some using large sophisticated computer programs. Once the plant is operational, the utility assumes the large responsibility for plant design. Utility personnel must understand the fundamentals of operating the plant, the technical information in the updated safety analysis report, all calculations used to design the plant, and the input for all design specification documents. Without this knowledge, utility personnel cannot successfully perform modifications or new analyses required by the NRC, such as probabilistic risk assessment (PRA) and motor-operated valve programs, and maintain the safe and reliable operation of the plant. Therefore, it is very important to have on-site personnel who understand how the calculations are performed and used in the design basis. This paper discusses the organization of the engineering analysis group, which provides technical support for River Bend Station (RBS) of Gulf States Utilities

Analysis on nuclear power plant control room system design and improvement based on human factor engineering

International Nuclear Information System (INIS)

Gao Feng; Liu Yanzi; Sun Yongbin

2014-01-01

The design of nuclear power plant control room system is a process of improvement with the implementation of human factor engineering theory and guidance. The method of implementation human factor engineering principles into the nuclear power plant control room system design and improvement was discussed in this paper. It is recommended that comprehensive address should be done from control room system function, human machine interface, digital procedure, control room layout and environment design based on the human factor engineering theory and experience. The main issues which should be paid more attention during the control room system design and improvement also were addressed in this paper, and then advices and notices for the design and improvement of the nuclear power plant control room system were afforded. (authors)

Refer to AP1000 for discussing the betterment of seismic design of internal nuclear power plant

International Nuclear Information System (INIS)

Gong Zhenbang; Zhang Renyan

2014-01-01

As a reference technique of AP1000, This paper discussed the betterment of seismic design of nuclear power plant in three ways. (1) Establish design criteria and guidelines for protection from seismic interaction; (2) Nuclear power plant seismic design of eliminating or weaken operation-basis earthquake; (3) Develop the seismic margin analysis (SMA) of the nuclear power plant. These three aspect are frontier technology in internal seismic design of internal nuclear power plant, and also these three technology are related intimately. (authors)

Demand analysis and design about NDPDMS

International Nuclear Information System (INIS)

Gao Yanfeng; Qian Bing; Wu Jianlei; Zhang Yishu; Shi Wenbin

1997-01-01

The authors give a simple introduction to Nuclear Power Engineering Database Plan and Development Management System (NDPDMS), including functions explanation, relations of subsystems, the interface of the system and other applications, etc. The system analysis, conceptual design and logic design are briefed

Generation IV nuclear plant design strategies

International Nuclear Information System (INIS)

Altin, V.

2007-01-01

In this presentation Generation IV nuclear reactor design criteria are examined under the light of known nuclear properties of fissile and fertile nuclei. Their conflicting nature is elucidated along with the resulting inevitability of a multitude of designs. The designs selected as candidates for further development are evaluated with respect to their potential to serve the different design criteria, thereby revealing their more difficult aspects of realization and the strong research challenges lying ahead

Development of Design Information Template for Nuclear Power Plants for Electromagnetic Pulse (EMP) Effect Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Minyi; Ryu, Hosan; Ye, Songhae; Lee, Euijong [KNHP CRI, Daejeon (Korea, Republic of)

2016-10-15

An electromagnetic pulse (EMP) is a transient electromagnetic shock wave that has powerful electric and magnetic fields that can destroy electronic equipment. It is generally well-known that EMPs can cause the malfunction and disorder of electronic equipment and serious damages to electric power systems and communication networks. Research is being carried out to protect nuclear power plants (NPPs) from EMP threats. Penetration routes of EMPs can be roughly categorized into two groups, radioactivity and conductivity. The radioactive effect refers to an impact transmitted to the ground from high-altitude electromagnetic pulses (HEMP). Such an impact may affect target equipment through the point of entry (POE) of the concrete structure of an NPP. The conductive effect refers to induced voltage or current coupled to the NPPs cable structure. The induced voltage and current affect the target equipment via connected cables. All these factors must be considered when taking into account EMP effect analysis for NPPs. To examine all factors, it is necessary to fully understand the schemes of NPPs. This paper presents a four type design information template that can be used to analyze the EMP effect in operating nuclear power plants. In order to analyze of the effects of EMPs on operating NPPs, we must consider both the conductive and radioactive effects on the target (system, equipment, structure). For these reasons, not only the equipment information, but also the information about the structure and the external penetration will be required. We are developing a design information template for robust nuclear design information acquisition. We expect to develop a block diagram on the basis of the template.

Development of Design Information Template for Nuclear Power Plants for Electromagnetic Pulse (EMP) Effect Analysis

International Nuclear Information System (INIS)

Kim, Minyi; Ryu, Hosan; Ye, Songhae; Lee, Euijong

2016-01-01

An electromagnetic pulse (EMP) is a transient electromagnetic shock wave that has powerful electric and magnetic fields that can destroy electronic equipment. It is generally well-known that EMPs can cause the malfunction and disorder of electronic equipment and serious damages to electric power systems and communication networks. Research is being carried out to protect nuclear power plants (NPPs) from EMP threats. Penetration routes of EMPs can be roughly categorized into two groups, radioactivity and conductivity. The radioactive effect refers to an impact transmitted to the ground from high-altitude electromagnetic pulses (HEMP). Such an impact may affect target equipment through the point of entry (POE) of the concrete structure of an NPP. The conductive effect refers to induced voltage or current coupled to the NPPs cable structure. The induced voltage and current affect the target equipment via connected cables. All these factors must be considered when taking into account EMP effect analysis for NPPs. To examine all factors, it is necessary to fully understand the schemes of NPPs. This paper presents a four type design information template that can be used to analyze the EMP effect in operating nuclear power plants. In order to analyze of the effects of EMPs on operating NPPs, we must consider both the conductive and radioactive effects on the target (system, equipment, structure). For these reasons, not only the equipment information, but also the information about the structure and the external penetration will be required. We are developing a design information template for robust nuclear design information acquisition. We expect to develop a block diagram on the basis of the template

A preliminary analysis of the BOP design management in nuclear power plant

International Nuclear Information System (INIS)

Tian Bin

2014-01-01

BOP project is an important part of nuclear power plant to maintain the normal operation and maintenance of the plant. The level of the design management has the important influences on the quality of the whole project, Design management includes the choice of the design standards, design evaluation, document control, the management of the design interface, design modification management and the management of the design service. The paper will start from various design issues in the construction of the Fuqing BOP project, analyse the causes of the problems on the design schedule management, interface, evaluation and the modification management. And then the paper also provides suggestions for improvement about all of this. (author)

Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

Directory of Open Access Journals (Sweden)

Ahmer Ali

2017-06-01

Full Text Available Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB under strong short-period ground motions (SPGMs and long-period ground motions (LPGMs. The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

Energy Technology Data Exchange (ETDEWEB)

Ali, Ahmer [ENVICO Consultants Co. Ltd., Seoul (Korea, Republic of); Abu-Hayah, Nadin; Kim, Doo Kie [Civil and Environmental Engineering, Kunsan National University, Gunsan (Korea, Republic of); Cho, Sung Gook [Innose Tech Co., Ltd., Incheon (Korea, Republic of)

2017-06-15

Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

Design concepts of nuclear desalination plants

International Nuclear Information System (INIS)

2002-11-01

Interest in using nuclear energy for producing potable water has been growing worldwide in the past decade. This has been motivated by a variety of factors, including economic competitiveness of nuclear energy, the growing need for worldwide energy supply diversification, the need to conserve limited supplies of fossil fuels, protecting the environment from greenhouse gas emissions, and potentially advantageous spin-off effects of nuclear technology for industrial development. Various studies, and at least one demonstration project, have been considered by Member States with the aim of assessing the feasibility of using nuclear energy for desalination applications under specific conditions. In order to facilitate information exchange on the subject area, the IAEA has been active for a number of years in compiling related technical publications. In 1999, an inter regional technical co-operation project on Integrated Nuclear Power and desalination System Design was launched to facilitate international collaboration for the joint development by technology holders and potential end users of an integrated nuclear desalination system. This publication presents material on the current status of nuclear desalination activities and preliminary design concepts of nuclear desalination plants, as made available to the IAEA by various Member States. It is aimed at planners, designers and potential end-users in those Member States interested in further assessment of nuclear desalination. Interested readers are also referred to two related and recent IAEA publications, which contain useful information in this area: Introduction of Nuclear Desalination: A Guidebook, Technical Report Series No. 400 (2000) and Safety Aspects of Nuclear Plants Coupled with Seawater Desalination Units, IAEA-TECDOC-1235 (2001)

Design of nuclear fuel cells by means of a statistical analysis and a sensibility study; Diseno de celdas de combustible nuclear mediante un analisis estadistico y un estudio de sensibilidad

Energy Technology Data Exchange (ETDEWEB)

Jauregui C, V. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Castillo M, J. A.; Ortiz S, J. J.; Montes T, J. L.; Perusquia del C, R., E-mail: alejandro.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

This work contains the results of the statistical analysis realized to study the nuclear fuel cells performance, considering the frequencies for the election of fuel bars used in the design of the same ones. The election of the bars used for the cells design are of 3 types, the first election shows that to the plotting the respective frequency is similar to a normal distribution, in the second case the frequencies graph is of type inverted square X{sup 2} and the last election is when the bars are chosen in aleatory form. The heuristic techniques used for the cells design were the neural networks, the ant colonies and a hybrid between the dispersed search and the trajectories re-linkage. To carry out the statistical analysis in the cells design were considered the local power peak factor and the neutron infinite multiplication factor (k∞) of this. On the other hand, the performance of the designed cells was analyzed when verifying the position of the bars containing gadolinium. The results show that is possible to design cells of nuclear fuel with a good performance, when considering the frequency of the bars used in their design. (Author)

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

Bimodal Nuclear Thermal Rocket Analysis Developments

Science.gov (United States)

Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

2014-01-01

Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

Design and analysis for piping systems

International Nuclear Information System (INIS)

Sterkel, H.-P.; Cutrim, J.H.C.

1981-01-01

The procedure and the typical techniques that are used in NUCLEN for the design and the calculation of the piping of Nuclear Plants. The classification system are generically described and the analysis techniques which are used for the design and verification of the piping systems, i.e. pressure design for the dimensioning of the wallthicknesses, temperature and dead weight analysis together with determination of support points, are shown. The techniques of dynamic design and analyses are described for earthquake and pressure impulse loadings. (Author) [pt

Economic Analysis of the Reference Design for a Nuclear-Driven High-Temperature-Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen consists of 4,009,177 cells with a per-cell active area of 225 cm2. A nominal cell area-specific resistance, ASR, value of 0.4 Ohm-cm2 with a current density of 0.25 A/cm2 was used, and isothermal boundary conditions were assumed. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current, AC, to direct current, DC, conversion is 96%. The overall system thermal-to-hydrogen production efficiency (based on the low heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of the plant was also performed using the H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost using realistic financial and cost estimating assumptions. A required cost of $3.23 per kg of hydrogen produced was calculated assuming an internal rate of return of 10%. Approximately 73% of this cost ($2.36/kg) is the result of capital costs associated with

AZTLAN platform: Mexican platform for analysis and design of nuclear reactors; AZTLAN platform: plataforma mexicana para el analisis y diseno de reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Gomez T, A. M.; Puente E, F. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edif. 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Francois L, J. L.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: armando.gomez@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

2014-10-15

The Aztlan platform Project is a national initiative led by the Instituto Nacional de Investigaciones Nucleares (ININ) which brings together the main public houses of higher studies in Mexico, such as: Instituto Politecnico Nacional, Universidad Nacional Autonoma de Mexico and Universidad Autonoma Metropolitana in an effort to take a significant step toward the calculation autonomy and analysis that seeks to place Mexico in the medium term in a competitive international level on software issues for analysis of nuclear reactors. This project aims to modernize, improve and integrate the neutron, thermal-hydraulic and thermo-mechanical codes, developed in Mexican institutions, within an integrated platform, developed and maintained by Mexican experts to benefit from the same institutions. This project is financed by the mixed fund SENER-CONACYT of Energy Sustain ability, and aims to strengthen substantially to research institutions, such as educational institutions contributing to the formation of highly qualified human resources in the area of analysis and design of nuclear reactors. As innovative part the project includes the creation of a user group, made up of members of the project institutions as well as the Comision Nacional de Seguridad Nuclear y Salvaguardias, Central Nucleoelectrica de Laguna Verde (CNLV), Secretaria de Energia (Mexico) and Karlsruhe Institute of Technology (Germany) among others. This user group will be responsible for using the software and provide feedback to the development equipment in order that progress meets the needs of the regulator and industry; in this case the CNLV. Finally, in order to bridge the gap between similar developments globally, they will make use of the latest super computing technology to speed up calculation times. This work intends to present to national nuclear community the project, so a description of the proposed methodology is given, as well as the goals and objectives to be pursued for the development of the

Optimum design of cogeneration system for nuclear seawater desalination - 15272

International Nuclear Information System (INIS)

Jung, Y.H.; Jeong, Y.H.

2015-01-01

A nuclear desalination process, which uses the energy released by nuclear fission, has less environmental impact and is generally cost-competitive with a fossil-fuel desalination process. A reference cogeneration system focused on in this study is the APR-1400 coupled with a MED (multi-effect distillation) process using the thermal vapor compression (TVC) technology. The thermal condition of the heat source is the most crucial factor that determines the desalination performance, i.e. energy consumption or freshwater production, of the MED-TVC process. The MED-TVC process operating at a higher motive steam pressure clearly shows a higher desalination performance. However, this increased performance does not necessarily translate to an advantage over processes operated at lower motive steam pressures. For instance, a higher motive steam pressure will increase the heat cost resulting from larger electricity generation loss, and thus may make this process unfavorable from an economic point of view. Therefore, there exists an optimum design point in the coupling configuration that makes the nuclear cogeneration system the most economical. This study is mainly aimed at investigating this optimum coupling design point of the reference nuclear cogeneration system using corresponding analysis tools. The following tools are used: MEE developed by the MEDRC for desalination performance analysis of the MED-TVC process, DE-TOP and DEEP developed by the IAEA for modeling of coupling configuration and economic evaluation of the nuclear cogeneration system, respectively. The results indicate that steam extraction from the MS exhaust and condensate return to HP FWHTR 5 is the most economical coupling design

A computational model for thermal fluid design analysis of nuclear thermal rockets

International Nuclear Information System (INIS)

Given, J.A.; Anghaie, S.

1997-01-01

A computational model for simulation and design analysis of nuclear thermal propulsion systems has been developed. The model simulates a full-topping expander cycle engine system and the thermofluid dynamics of the core coolant flow, accounting for the real gas properties of the hydrogen propellant/coolant throughout the system. Core thermofluid studies reveal that near-wall heat transfer models currently available may not be applicable to conditions encountered within some nuclear rocket cores. Additionally, the possibility of a core thermal fluid instability at low mass fluxes and the effects of the core power distribution are investigated. Results indicate that for tubular core coolant channels, thermal fluid instability is not an issue within the possible range of operating conditions in these systems. Findings also show the advantages of having a nonflat centrally peaking axial core power profile from a fluid dynamic standpoint. The effects of rocket operating conditions on system performance are also investigated. Results show that high temperature and low pressure operation is limited by core structural considerations, while low temperature and high pressure operation is limited by system performance constraints. The utility of these programs for finding these operational limits, optimum operating conditions, and thermal fluid effects is demonstrated

Sealing analysis for nuclear vessel of PWR

International Nuclear Information System (INIS)

Qu, J.; Dou, Y.

1987-01-01

Although design by analysis of pressure vessel has become a requirement in all codes for more than 20 years, sealing design for nuclear components is still too complicated and there are yet no criteria about this aspect, even though in the well-known ASME Boiler and Pressure Vessel Code. Thus it is of significance to undertake researches of transient sealing tests and analysis for nuclear vessel. Since 1960s great progress has been made in analytic computer program, which takes flange as a rigid ring. Actually, however, there are elastic or elastoplastic contacts on flange mating surface. Chen (1979) gave a mixed finite element method, using a condensing flexible matrix skill, to solve two-body contact problem. On the basis of axisymmetric stress and thermal analysis of finite element method and on accepting Chen's (1979) idea of mixed finite element method, we have developed a computer program for sealing analysis, named SMEC, which considers bolt loading changes and temperature effects. (orig./GL)

Protection against internal fires and explosions in the design of nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Experience of the past two decades in the operation of nuclear power plants and modern analysis techniques confirm that fire may be a real threat to nuclear safety and should receive adequate attention from the beginning of the design process throughout the life of the plant. Within the framework of the NUSS programme, a Safety Guide on fire protection had therefore been developed to enlarge on the general requirements given in the Code. Since its first publication in 1979, there has been considerable development in protection technology and analysis methods and after the Chernobyl accident it was decided to revise the existing Guide. This Safety Guide supplements the requirements established in Safety of Nuclear Power Plants: Design. It supersedes Safety Series No. 50-SG-D2 (Rev. 1), Fire Protection in Nuclear Power Plants: A Safety Guide, issued in 1992.The present Safety Guide is intended to advise designers, safety assessors and regulators on the concept of fire protection in the design of nuclear power plants and on recommended ways of implementing the concept in some detail in practice

Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

2016-05-15

The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

Facility Safeguardability Analysis in Support of Safeguards by Design

International Nuclear Information System (INIS)

Wonder, E.F.

2010-01-01

The idea of 'Safeguards-by-Design' (SBD) means designing and incorporating safeguards features into new civil nuclear facilities at the earliest stages in the design process to ensure that the constructed facility is 'safeguardable,' i.e. will meet national and international nuclear safeguards requirements. Earlier consideration of safeguards features has the potential to reduce the need for costly retrofits of the facility and can result in a more efficient and effective safeguards design. A 'Facility Safeguardability Analysis' (FSA) would be a key step in Safeguards-by-Design that would link the safeguards requirements with the 'best practices', 'lessons learned', and design of the safeguards measures for implementing those requirements. The facility designer's nuclear safeguards experts would work closely with other elements of the project design team in performing FSA. The resultant analysis would support discussions and interactions with the national nuclear regulator (i.e. State System of Accounting for and Control of Nuclear Material - SSAC) and the IAEA for development and approval of the proposed safeguards system. FSA would also support the implementation of international safeguards by the IAEA, by providing them with a means to analyse and evaluate the safeguardability of facilities being designed and constructed - i.e. by independently reviewing and validating the FSA as performed by the design team. Development of an FSA methodology is part of a broader U.S. National Nuclear Security Administration program to develop international safeguards-by-design tools and guidance documents for use by facility designers. The NNSA NGSI -sponsored project team is looking, as one element of its work, at how elements of the methodology developed by the Generation IV International Forum's Working Group on Proliferation Resistance and Physical Protection can be adapted to supporting FSA. (author)

Development of an evaluation method for seismic isolation systems of nuclear power facilities. Seismic design analysis methods for crossover piping system

International Nuclear Information System (INIS)

Tai, Koichi; Sasajima, Keisuke; Fukushima, Shunsuke; Takamura, Noriyuki; Onishi, Shigenobu

2014-01-01

This paper provides seismic design analysis methods suitable for crossover piping system, which connects between seismic isolated building and non-isolated building in the seismic isolated nuclear power plant. Through the numerical study focused on the main steam crossover piping system, seismic response spectrum analysis applying ISM (Independent Support Motion) method with SRSS combination or CCFS (Cross-oscillator, Cross-Floor response Spectrum) method has found to be quite effective for the seismic design of multiply supported crossover piping system. (author)

Design of compact nuclear power marine engineering simulator

International Nuclear Information System (INIS)

Gao Jinghui; Xing Hongchuan; Zhang Ronghua; Yang Yanhua; Xu Jijun

2004-01-01

The essentiality of compact nuclear power marine engineering simulator (NPMES) is discussed. The technology of nuclear power plant engineering simulator (NPPES) for NPMES development is introduced, and the function design, general design and model design are given in details. A compact NPMES based on the nuclear power marine of 'Mutsu' is developed. The design can help the development of NPMES, which will improve operation safety and management efficiency of marine. (authors)

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

International Nuclear Information System (INIS)

1983-04-01

The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

Salt creep design consideration for underground nuclear waste storage

International Nuclear Information System (INIS)

Li, W.T.; Wu, C.L.; Antonas, N.J.

1983-01-01

This paper summarizes the creep consideration in the design of nuclear waste storage facilities in salt, describes the non-linear analysis method for evaluating the design adequacy, and presents computational results for the current storage design. The application of rock mechanics instrumentation to assure the appropriateness of the design is discussed. It also describes the design evolution of such a facility, starting from the conceptual design, through the preliminary design, to the detailed design stage. The empirical design method, laboratory tests and numerical analyses, and the underground in situ tests have been incorporated in the design process to assure the stability of the underground openings, retrievability of waste during the operation phase and encapsulation of waste after decommissioning

Seismic analysis, evaluation and upgrade design for a nuclear facility exhaust stack building

International Nuclear Information System (INIS)

Malik, L.E.; Kabir, A.F.

1991-01-01

This paper reports on an exhaust stack building of a nuclear reactor facility with complex structural configuration that has been analyzed and evaluated for seismic forces. This building was built in the 1950's and had not been designed to resist seismic forces. A very rigorous analysis and evaluation program was implemented to minimize the costly retrofits required to upgrade the building to resist high seismic forces. The seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, base mat flexibility and the influence of the nearby reactor building have been considered in the seismic analyses. The rigorous analyses and evaluation enabled limited upgrades to qualify the stack building for the seismic forces

Inelastic design of nuclear reactor structures and its implications on design of critical equipment

International Nuclear Information System (INIS)

Newmark, N.M.

1977-01-01

In considering the response of a nuclear reactor structure to seismic motions, one must take account of the implications of various levels of damage, short of impairment of safety, and definitely short of collapse, of the structure. Some structural elements of nuclear power plants must perforce remain elastic or nearly elastic in order to perform their allocated safety function. However, in many instances, a purely linear elastic analysis may be unreasonably conservative when one considers that even up to the near yield point range, there are nonlinearities of sufficient amount to reduce required design levels considerably. Moreover, limited yielding of a structure may reduce the response of equipment located in the structure below those levels of response that would be excited were the structure to remain elastic. Energy absorption in the inelastic range is most conveniently treated by use of the so-called 'ductility factor' introduced by the author for design of structures and equipment to resist explosion and blast forces. In general, for small excursions into the inelastic range, especially when the latter can be approximated by an elasto-plastic resistance curve, the design response spectrum is decreased by a simply determined factor that is related to the ductility factor. Many important parts of equipment of a nuclear power plant facility are attached to the principal parts of the structure and respond in a manner determined by the structural response as well as by the general ground motion to which the structure is subjected. This matter involves some difficulty in analysis, but appropriate calculational techniques and design methods are available. A suitable design simplification is one in which the response of the attachment is related to the modal responses of the structure. This equipment response is affected by the relative mass of the attachment and the structure

Application of analysis technology in nuclear plant

International Nuclear Information System (INIS)

Takaoka, Keiko; Miura, Hiromi; Umeda, Kenji

1996-01-01

Recently, thanks to the rapid improvement of EWS performance, the authors have been able to carry out design evaluation comparatively, easily, utilizing computational fluid dynamics (CFD) technology. The Nuclear Plant Engineering Department has carried out some analyses in the past several years with the main purpose of evaluating the design of nuclear reactor internals. These studies included ''Thermal Hydraulic Analysis for Top Plenum'' and ''Flow Analysis for Lower Plenum''. It is considered to be a special matter in thermal hydraulic analysis of the top plenum that temperature distribution has been estimated with a relatively small number of meshes by means of an imaginary spray nozzle, and in the flow analysis for the lower plenum that flow distribution has been found to change largely, depending on the reactor internals. One of the ways to confirm the safety of nuclear plants, detailed structural analysis, is required for all possible combinations of transient and load conditions during operation. In particular, it is very important to clarify the thermal stress behavior under operating conditions and to evaluate fatigue analysis in accordance with the Code Requirements. However, it is very complicated and it takes a lot of time. A new system was developed which can operate continuously all of the definitions of the analytical model, the analyzation of pressurized thermal and external stress, and editing reports. In this paper, the authors introduce this system and apply it to a pressurized water reactor

Cognitive task analysis of nuclear power plant operators for man-machine interface design

International Nuclear Information System (INIS)

Itoh, J.I.; Yoshimura, S.; Ohtsuka, T.

1990-01-01

This paper aims to ascertain and further develop design guidelines for a man-machine interface compatible with plant operators' problem solving strategies. As the framework for this study, operator's information processing activities were modeled, based on J. Rasmussen's framework for cognitive task analysis. Two experiments were carried out. One was an experiment aimed at gaining an understanding of internal mechanisms involved in mistakes and slips which occurred in operators' responses to incidents and accidents. As a result of fifteen cases of operator performance analysis, sixty one human errors were identified. Further analysis of the errors showed that frequently occurring error mechanisms were absent-mindedness, lack of recognition of patterns in diagnosis and failed procedure formulation due to memory lapses. The other kind of experiment was carried out to identify the envelope of trajectories for the operator's search in the problem space consisting of the two dimensions of means-ends and whole-part relations while dealing with transients. Two cases of experimental sessions were conducted with the thinking-aloud method. From analyses based on verbal protocols, trajectories of operator's search were derived, covering from the whole plant level through the component level in the whole-part dimension and covering from the functional purpose level through the physical form level in the means-ends dimension. The findings obtained from these analyses serve as a basis for developing design guidelines for man-machine interfaces in control rooms of nuclear power plants

International certification of nuclear power reactors design. A proposal from the U.S. NRC (Nuclear Regulatory Commission)

International Nuclear Information System (INIS)

Felizia, Eduardo R.

2006-01-01

The proposal foundations of the Nuclear Regulatory Commission Board Chairman are briefly described, which were enunciated at a meeting on Fourth Generation Reactors (Washington, March 2005). This proposal is analyzed mainly from the point of view of its consequences in third countries buyers of nuclear technology. The analysis is complemented by descriptions of the current process of the NRC design certification and of Third and Fourth Generation Reactors. (author) [es

Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

International Nuclear Information System (INIS)

Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

2000-09-01

A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

{sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

Design assurance programme for modifications at the Ringhals Nuclear Power Plant

International Nuclear Information System (INIS)

Persson, K.

1985-01-01

The Engineering Department within the site organization is responsible for modifications and improvements of plant design. Internal and external specialists are also involved in the design activities. Modifications follow the procedures laid down in the Design Assurance Programme (DAP), according to the overall directions given in 10CFR50 APP B and the more detailed recommendations in American Nuclear Standard 3.2, and the United States Nuclear Regulatory Commission's Regulatory Guides 1.33 and 1.64. The site Quality Assurance Department has adapted these directions to Swedish conditions. All plant modifications are ordered from customers within the site organization. The design work is performed according to the following steps: priority of problems or requirements; overall analysis of problems or requirements; basic design (preliminary and final); detailed design (electrical, mechanical, civil); purchase of equipment; installation in plant; testing of installed equipment. The project engineer is responsible for modifications from the initiation of analysis until testing and documentation are finalized. For each step, procedures and checklists are available. Internal and external specialists as well as the Swedish Nuclear Inspectorate are involved when appropriate. Design engineers and QC/QE engineers are also involved. Special groups within the site organization handle purchase and installation. Testing the installed equipment is performed in co-operation with the operation staff. Deviations from DAP must be reported by all personnel, and corrective actions must be taken. Internal audits have to be done once a year for the entire DAP and the audit, with its findings and the corrective actions taken, must be reported to the QA group. (author)

The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

NARCIS (Netherlands)

Sjenitzer, B.L.

2013-01-01

In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing

GE's advanced nuclear reactor designs

International Nuclear Information System (INIS)

Berglund, R.C.

1993-01-01

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

Probabilistic design of nuclear structures: a summary of state of the art and research needs

International Nuclear Information System (INIS)

Ravindra, M.K.; Walser, A.

1978-01-01

This paper provides an overview of ongoing research in probabilistic design of nuclear structures. The main areas of review are (1) loads, (2) load combinations, (3) missiles, (4) design criteria, (5) seismic safety, (6) system reliability, (7) hazard analysis, and (8) probabilistic response. A consistent framework of probabilistic design of nuclear structures is proposed. Areas of further research and data collection are suggested. (Auth.)

Nuclear reactor descriptions for space power systems analysis

Science.gov (United States)

Mccauley, E. W.; Brown, N. J.

1972-01-01

For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

Design basis reconstitution and configuration management of nuclear power plants

International Nuclear Information System (INIS)

Smith, P.R.

1989-01-01

The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems

Aspects of the design and structural analysis of the prestressed cast iron nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Thomas, R.G.

1978-09-01

The development of the prestressed cast iron nuclear reactor pressure vessel up to the present time is reviewed, and the current status is outlined of the techniques used for its structural analysis. Details of the manufacturing processes involved in the production of the castings, and problems of inspecting them to the standards required for a nuclear application are discussed. A method for the detailed modelling of the cast iron segments is proposed, using the finite element technique with plate bending elements, and criteria for obtaining accurate results are derived. The application of the technique to the analysis of a single cast segment situated in the wall of a PCIPV has enabled an accurate determination of the stress field to be made. Account is taken of the effect of the vessel displacements on the tendon stresses at normal vault pressure and at high overpressure. Studies by this method of several different casting designs have identified favourable features, which have been incorporated into an optimised design. The sensitivity of the structure to a machining error in a casting and to the failure or removal of circumferential and axial tendons is examined, making use of axisymmetric and three-dimensional global finite element solutions to provide boundary conditions for detailed local analyses. Some aspects of the economics of the cast iron reactor pressure vessel are discussed, and recommendations are made for further research in areas relevant to the assessment of the reliability of the vessel. (author)

Trend analysis in the nuclear maintenance industry

International Nuclear Information System (INIS)

Ruemeli, W.A.

1986-01-01

The maintenance of nuclear facilities is a demanding, ongoing activity which requires the same level of quality as new construction. Heretofore, many owners and contractors have relied on ''gut feel'' to determine whether maintenance quality was improving or not. However, trend analysis now is becoming a key factor in monitoring plant activities to ensure quality. Literature abounds with descriptions of computerized systems for collecting and sorting data. Even the Nuclear Regulatory Commission (NRC) has concurred, with its endorsement of trend analysis of construction indicators in NUREG 1055 (Ford Amendment Study). Stearns Catalytic has developed a unique system of tend analyses for nuclear plant activities. Aside from its intended purpose of determining the quality trends in maintenance activities, the program also supplies substantial quantitative data for control and management of the quality activities. Trend analysis is a time series analysis of a set of observations arranged in chronological order. The important aspect is the time basis, specifically the analysis of quality indicators over successive periods of time. Many program elements, including surveillances, nonconformances, inspections, and audits, are designed to look at quality indications

The computer code system for reactor radiation shielding in design of nuclear power plant

International Nuclear Information System (INIS)

Li Chunhuai; Fu Shouxin; Liu Guilian

1995-01-01

The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

Structure design of water discharge surge tank of nuclear power plant

International Nuclear Information System (INIS)

Wang Fang; Hou Shuqiang

2015-01-01

Drainage is an important function of water discharge surge tank in nuclear power plant. There is little wall and beam inside the water discharge surge tank due to the requirement of major work, which is different from the general structure. Taking water discharge surge tank of nuclear power plant for example, concerned problems are expatiated in the structure scheme of water discharge surge tank, and important structural components are analyzed. Structural analysis model is established by ANSYS finite element analysis. A comprehensive and numerical analysis is performed for different combinations of structural model, and the internal force of structure is extracted. Finally, suggestions for design of similar structure are proposed. (authors)

Study on integrated design and analysis platform of NPP

International Nuclear Information System (INIS)

Lu Dongsen; Gao Zuying; Zhou Zhiwei

2001-01-01

Many calculation software have been developed to nuclear system's design and safety analysis, such as structure design software, fuel design and manage software, thermal hydraulic analysis software, severe accident simulation software, etc. This study integrates those software to a platform, develops visual modeling tool for Retran, NGFM90. And in this platform, a distribution calculation method is also provided for couple calculation between different software. The study will improve the design and analysis of NPP

Verification and uncertainty evaluation of CASMO-3/MASTER nuclear analysis system

Energy Technology Data Exchange (ETDEWEB)

Song, Jae Seung; Cho, Byung Oh; Joo, Han Kyu; Zee, Sung Quun; Lee, Chung Chan; Park, Sang Yoon

2000-06-01

MASTER is a nuclear design code developed by KAERI. It uses group constants generated by CASMO-3 developed by Studsvik. In this report the verification and evaluation of uncertainty were performed for the code system application in nuclear reactor core analysis and design. The verification is performed via various benchmark comparisons for static and transient core condition, and core follow calculations with startup physics test predictions of total 14 cycles of pressurized water reactors. Benchmark calculation include comparisons with reference solutions of IAEA and OECA/NEA problems and critical experiment measurements. The uncertainty evaluation is focused to safety related parameters such as power distribution, reactivity coefficients, control rod worth and core reactivity. It is concluded that CASMO-3/MASTER can be applied for PWR core nuclear analysis and design without any bias factors. Also, it is verified that the system can be applied for SMART core, via supplemental comparisons with reference calculations by MCNP which is a probabilistic nuclear calculation code.

Verification and uncertainty evaluation of CASMO-3/MASTER nuclear analysis system

International Nuclear Information System (INIS)

Song, Jae Seung; Cho, Byung Oh; Joo, Han Kyu; Zee, Sung Quun; Lee, Chung Chan; Park, Sang Yoon

2000-06-01

MASTER is a nuclear design code developed by KAERI. It uses group constants generated by CASMO-3 developed by Studsvik. In this report the verification and evaluation of uncertainty were performed for the code system application in nuclear reactor core analysis and design. The verification is performed via various benchmark comparisons for static and transient core condition, and core follow calculations with startup physics test predictions of total 14 cycles of pressurized water reactors. Benchmark calculation include comparisons with reference solutions of IAEA and OECA/NEA problems and critical experiment measurements. The uncertainty evaluation is focused to safety related parameters such as power distribution, reactivity coefficients, control rod worth and core reactivity. It is concluded that CASMO-3/MASTER can be applied for PWR core nuclear analysis and design without any bias factors. Also, it is verified that the system can be applied for SMART core, via supplemental comparisons with reference calculations by MCNP which is a probabilistic nuclear calculation code

Design aspects of radiological safety in nuclear facilities

International Nuclear Information System (INIS)

Patkulkar, D.S.; Purohit, R.G.; Tripathi, R.M.

2014-01-01

In order to keep operational performance of a nuclear facility high and to keep occupational and public exposure ALARA, radiological safety provisions must be reviewed at the time of facility design. Deficiency in design culminates in deteriorated system performance and non adherence to safety standards and could sometimes result in radiological incident. Important radiological aspects relevant to safety were compiled based on operating experiences, design deficiencies brought out from past nuclear incidents, experience gained during maintenance, participation in design review of upcoming nuclear facilities and radiological emergency preparedness

Investigation of human system interface design in nuclear power plant

International Nuclear Information System (INIS)

Feng Yan; Zhang Yunbo; Wang Zhongqiu

2012-01-01

The paper introduces the importance of HFE in designing nuclear power plant, and introduces briefly the content and scope of HFE, discusses human system interface design of new built nuclear power plants. This paper also describes human system interface design of foreign nuclear power plant, and describes in detail human system interface design of domestic nuclear power plant. (authors)

Consequence analysis for nuclear reactors, Yongbyon

International Nuclear Information System (INIS)

Kang, Taewook; Jae, Moosung

2017-01-01

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)

Ship nuclear power plants: system approach designing and optimal parameter selection

International Nuclear Information System (INIS)

Dolgov, V.N.

1995-01-01

Dependence of basic indices of naval nuclear power installations on combination of thermotechnical parameters is considered. The necessity of new approach to designing naval nuclear power facilities, based on combined analysis of the above facilities and the ship, as a single complex system, is demonstrated. The conclusion is made that unjustified increase in the ship metal capacity, terms and costs of their production takes place in case of disregarding the role of mass and dimensions of nuclear power facilities and consequently their thermotechnical characteristics in formation of the ship displacement tonnage. 2 refs., 3 figs

Comparison of design and probabilistic analyses of nuclear power plants

International Nuclear Information System (INIS)

Maslenikov, O.R.; Johnson, J.J.; Campbell, R.D.

1995-01-01

A study was made to evaluate the margin of conservatism introduced into design in-structures response spectra by following standard design analysis procedures according to the U.S.Nuclear Regulatory Commission (NRC) Standard Review Plan and Regulatory Guides for comparing spectra produced by such a design analysis to response from median-centered probabilistic analyses. Three typical nuclear plant structures were studied: PWR reactor building, PWR auxiliary building and BWR reactor building. Each building was assumed to be situated on three idealized sites: a rock site, a medium and a soft soil site. All buildings were assumed to have embedded foundations. The PWR reactor building was also assumed to have a surface foundation. Each design analysis was performed inn accordance with the current SRP criteria. Each probabilistic analysis consisted of 30 earthquake simulations for which the free-field motions and soil and structural properties were varied; the simulated earthquakes were generated such that their mean-plus-one-standard-deviation free-field spectra approximated the Regulatory Guide (RG) 1.60 design spectra. In-structure response spectra from the design analyses were compared with the 84% non-exceedance probability (NEP) spectra from the probabilistic analyses. The comparisons showed that the design method produced conservative results for all cases. The smallest margin was about 10% for buildings on rock sites. Softer sides had larger margins of conservatism; the reactor buildings on the soft soil site had margins of as much as 100% (factor of 2),. The shorter structures and lower locations in all buildings had smaller margins. The margin of conservatism for the surface founded reactor building was about 20% more than for the embedded reactor building. (author). 3 refs., 5 figs., 1 tab

Nuclear safety and radiation protection consideration in the design of research and development facility

International Nuclear Information System (INIS)

Akbar, M.R.

2010-01-01

Nuclear safety is a critically important aspect that must be considered in the design of a nuclear facility in order to ensure the protection of the workers, public and environment. This paper looks at the methodology, approach and incorporation of this aspect, specifically into the design of a research and development facility. The Health, Safety and Environmental Basis of Design is an initial analysis of nuclear safety and radiation protection considerations that is performed during the conceptual design phase and sets the baseline for what the design of the facility must conform to. It consists of general nuclear safety design principles, such as defence in depth and optimisation considerations, and a hazard management strategy. Following the Health, Safety and Environmental Basis of Design, a Preliminary Safety Assessment Report is generated during the basic design phase in conjunction with various analyses in order to assess the impact of hazards on the workers and members of the public. This assessment follows a hazard graded approach where the depth of the analysis will be determined by the impact of the worst case accident scenario in the facility. The assessment also includes a waste management strategy which is an essential aspect to be considered in the design in order to minimize the generation of waste. The safety assessment also demonstrates compliance to dose limits and risk criteria for the workers and members of the public set by the regulatory body and supported by a legal framework. Measures are taken to keep risk as low as reasonably achievable and prevent transgression of the risk and dose limits. However, a balance needs to be maintained between 5 reducing these doses further and the cost of such a reduction, which is known as optimization. It is therefore imperative to have nuclear safety specialists analyse the design in order to protect the worker and member of the public from unwarranted exposure to nuclear radiation. (author)

Quality assurance of analytical, scientific, and design computer programs for nuclear power plants

International Nuclear Information System (INIS)

1994-06-01

This Standard applies to the design and development, modification, documentation, execution, and configuration management of computer programs used to perform analytical, scientific, and design computations during the design and analysis of safety-related nuclear power plant equipment, systems, structures, and components as identified by the owner. 2 figs

Quality assurance of analytical, scientific, and design computer programs for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-06-01

This Standard applies to the design and development, modification, documentation, execution, and configuration management of computer programs used to perform analytical, scientific, and design computations during the design and analysis of safety-related nuclear power plant equipment, systems, structures, and components as identified by the owner. 2 figs.

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

Magnet Design Considerations for Fusion Nuclear Science Facility

Energy Technology Data Exchange (ETDEWEB)

Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kessel, C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States) Fusion Technology Institute; Titus, P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.

Nuclear data for designing the IFMIF accelerator

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

The objective of the International Fusion Materials Irradiation Facility (IFMIF) and the design concept of the IFMIF accelerator system are described. The status of the nuclear data, especially for the deuteron-induced reactions, to qualify the system design is reviewed. The requests for the nuclear data compilation and/or evaluation are summarized. (author)

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

Ground motion prediction needs for nuclear engineering design

International Nuclear Information System (INIS)

Hadjian, A.H.

1985-01-01

The basic design philosophy of nuclear power plants stipulates that the risk to the public be as low as reasonably achievable. As a result of this philosophy, the seismic design of nuclear power plants has tended, over time, to diverge from that of other engineered structures. The emphasis at the present time is to specify ground motion at a nuclear facility site as realistically as possible and to design all safety-related structures to respond to the specified ground motion in the elastic range. The characteristics of this realistic design ground motion are discussed and present prediction needs identified

Seismic design of nuclear power plants. Where are we now?

International Nuclear Information System (INIS)

Roesset, J.M.

1995-01-01

The lack of any significant activity in the design and construction of new nuclear power plants over the last ten years has resulted in a corresponding lull in the basic academic research carried out in this field. While some work is still going on related to the evaluation of existing plants or to litigation over some of them (including some that never became operational) most of it is of a very applied nature and little basic research is being conducted at present. Yet research on earthquake engineering in general, as applied to buildings, bridges, lifelines, dams and other constructed facilities has continued. This paper attempts to look at some of the areas where there were major uncertainties in the seismic design of nuclear power plants (selection of the design earthquake and its characteristics, evaluation of soil effects and soil structure interactions, dynamic analysis and design of the structures), the progress that has been made in these areas, and the remaining issues in need of further research. (author)

Seismic design of nuclear power plants - where are we now?

International Nuclear Information System (INIS)

Roesset, J.M.

1998-01-01

The lack of any significant activity in the design and construction of new nuclear power plants over the last 10 years has resulted in a corresponding lull in the basic academic research carried out in this field. Whilst some work is still going on related to the evaluation of existing plants or to litigation over some of them (including some that never became operational) most of it is of a very applied nature and little basic research is being conducted at present. However, research on earthquake engineering in general, as applied to buildings, bridges, lifelines, dams and other constructed facilities has continued. This paper attempts to look at some of the areas where there were major uncertainties in the seismic design of nuclear power plants (selection of the design earthquake and its characteristics, evaluation of soil effects and soil structure interactions, dynamic analysis and design of the structures), the progress that has been made in these areas, and the remaining issues in need of further research. (orig.)

A risk-informed perspective on deterministic safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Wan, P.T.

2009-01-01

In this work, the deterministic safety analysis (DSA) approach to nuclear safety is examined from a risk-informed perspective. One objective of safety analysis of a nuclear power plant is to demonstrate via analysis that the risks to the public from events or accidents that are within the design basis of the power plant are within acceptable levels with a high degree of assurance. This nuclear safety analysis objective can be translated into two requirements on the risk estimates of design basis events or accidents: the nominal risk estimate to the public must be shown to be within acceptable levels, and the uncertainty in the risk estimates must be shown to be small on an absolute or relative basis. The DSA approach combined with the defense-in-depth (DID) principle is a simplified safety analysis approach that attempts to achieve the above safety analysis objective in the face of potentially large uncertainties in the risk estimates of a nuclear power plant by treating the various uncertainty contributors using a stylized conservative binary (yes-no) approach, and applying multiple overlapping physical barriers and defense levels to protect against the release of radioactivity from the reactor. It is shown that by focusing on the consequence aspect of risk, the previous two nuclear safety analysis requirements on risk can be satisfied with the DSA-DID approach to nuclear safety. It is also shown the use of multiple overlapping physical barriers and defense levels in the traditional DSA-DID approach to nuclear safety is risk-informed in the sense that it provides a consistently high level of confidence in the validity of the safety analysis results for various design basis events or accidents with a wide range of frequency of occurrence. It is hoped that by providing a linkage between the consequence analysis approach in DSA with a risk-informed perspective, greater understanding of the limitation and capability of the DSA approach is obtained. (author)

Optimization design for drain to nuclear power condenser

International Nuclear Information System (INIS)

Ding Jiapeng; Jiang Chengren

2010-01-01

Characters and varieties of drain to nuclear power condenser are discussed in this paper. Take the main steam system of a nuclear power as an example, normal and detailed optimization design are introduced, related expatiate are used as a reference for the drain of other systems. According to the characters of nuclear power instant operation, the influence and needed actions related with the optimization design are also analyzed. Based on the above research, the scheme has been carried out in a nuclear power station and safety for the condenser operation of the nuclear power has been improved largely. (authors)

Research needs and improvement of standards for nuclear power plant design

International Nuclear Information System (INIS)

Chen, C.; Moreadith, F.L.

1978-01-01

The need for research and improvement of code requirements, for both economy and safety reasons is discussed for the following topics relevant to nuclear power plant structural analysis: Earthquake definition; dynamic behavior of reinforced concrete structures under impact loads; design for postulated pipe rupture; code requirements for loading combinations for concrete structures, reinforcing steel splicing, reinforced concrete structural design for thermal effects. (Auth.)

Innovative designs and technologies of nuclear power. IV International scientific and technical conference. Book of abstracts

International Nuclear Information System (INIS)

2016-01-01

IV International scientific and technical conference “Innovative designs and technologies of nuclear power” has been organized and is conducted by JSC NIKIET with support from Rosatom State Corporation, the International Atomic Energy Agency, the Russian Academy of Sciences and the Nuclear Society of Russia. The conference topics include: innovative designs of nuclear facilities for various applications, nuclear fuel and new materials, closed fuel cycle technologies, SNF and RW management, technological answers to nonproliferation problems, small power reactors (stationary, transportable, floatable, propulsion, space), integrated codes of a new generation for safety analysis of nuclear power plants and fuel cycles, controlled fusion [ru

Subchannel analysis in nuclear reactors

International Nuclear Information System (INIS)

Ninokata, H.; Aritomi, M.

1992-01-01

This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

Rules for design of nuclear graphite core components - some considerations and approaches

International Nuclear Information System (INIS)

Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

1978-01-01

The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

Analysis of French (Paluel) pressurized water reactor design differences compared to current US PWR designs

International Nuclear Information System (INIS)

1986-05-01

To understand better the regulatory approaches to reactor safety in foreign countries, the staff of the Nuclear Regulatory Commisssion has reviewed design information on the Paluel nuclear power plant, one of the current standard 1300-MWe plant operating in France. This report provides the staff's evaluation of major design differences between this standardized French plant and current US pressurized water reactor plants, as well as insights concerning French regulatory practices. The staff identified approximately 25 design differences, and an analysis of the safety significance of each of these design features is presented, along with an assessment comparing the relative safety benefit of each

Nuclear power plant equipment design and construction rules

International Nuclear Information System (INIS)

Boiron, P.

1983-03-01

Presentation of the AFCEN (French association for nuclear power plant equipment design and construction rules) working, of its edition activity and of somes of its edited documents such as RCC-C (design and construction rules for PWR power plant fuel assemblies) and RCC-E (design and construction rules for nuclear facility electrical equipments) [fr

Seismic design criteria for nuclear powerplants

International Nuclear Information System (INIS)

Jennings, P.C.; Guzman, R.A.

1975-01-01

There are three main aspects of the problem of selection of seismic design criteria for major projects such as nuclear power plants. These are the description of the appropriate level of shaking to be considered, usually given in the form of design spectra; the allowable response of the structure, usually specified in terms of allowable stresses and deflections; and the capability of the structure to dissipate energy, commonly given in the form of fractions of critical damping. In this presentation only the first of these features is examined, with particular application to nuclear power plants. Under these restrictions, the most important parts of the problem become the determination of the amplitude of the design spectra corresponding to the safe shutdown earthquake (SSE) and the question of whether the shape of the spectra recommended by Regulatory Guide 1.60 (U. S. Atomic Energy Commission, 1973) is appropriate for the particular application. In the course of working out the details of the approach, it was found useful to reexamine a number of concepts including the use of response spectra or peak values of ground motion parameters, the shape of the design spectra, problems in attenuation and scaling, and the use of motions on the ground surface or bedrock motions. There is nothing fundamentally new in the suggested approach, although some of the features may not have been applied to the problem of selecting design spectra for nuclear power plants in the way suggested. The approach is applied only to nuclear power plants but it is not limited to this application

Westinghouse Nuclear Core Design Training Center - a design simulator

International Nuclear Information System (INIS)

Altomare, S.; Pritchett, J.; Altman, D.

1992-01-01

The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

Designing nuclear power plants for improved operation and maintenance

International Nuclear Information System (INIS)

1996-09-01

The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig

Designing nuclear power plants for improved operation and maintenance

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig.

Optimal design of base isolation and energy dissipation system for nuclear power plant structures

International Nuclear Information System (INIS)

Zhou Fulin

1991-01-01

This paper suggests the method of optimal design of base isolation and energy dissipation system for earthquake resistant nuclear power plant structures. This method is based on dynamic analysis, shaking table tests for a 1/4 scale model, and a great number of low cycle fatigue failure tests for energy dissipating elements. A set of calculation formulas for optimal design of structures with base isolation and energy dissipation system were introduced, which are able to be used in engineering design for earthquake resistant nuclear power plant structures or other kinds of structures. (author)

Safeguards by Design - Experiences from New Nuclear Installation

International Nuclear Information System (INIS)

Okko, O.; Honkamaa, T.; Kuusi, A.; Rautjaervi, J.

2010-01-01

The experiences obtained from the current construction projects at Olkiluoto clearly point out the need to introduce the safeguards requirements into facility design process at an early stage. The early Design Information is completed, in principle, before the construction. However, during the design of containment, surveillance systems, and non-destructive assay equipment and their cabling, the design requirements for safeguards systems were not available either for the new reactor unit or for the disposal plant with a geological repository. Typically, the official Design Information documents are not available early enough for efficient integration of safeguards systems into new facilities. In case of the Olkiluoto projects, this was due to understandable reasons: at the new reactor unit the design acceptance by the ordering company and by the nuclear safety authorities was a long process, ongoing simultaneously with parts of the construction; and at the geological repository the national legislation assigns the repository the status of a nuclear facility only after the initial construction and research phase of the repository when the long-term safety of the disposal concept is demonstrated. As similar factors are likely to delay the completion of the official Design Information documents with any new reactor projects until the construction is well underway and efficient integration of safeguards systems is impossible. Therefore, the proliferation resistance of new nuclear installations should be addressed in the design phase before the official Design Information documents are finished. This approach was demonstrated with the enlargement of the Olkiluoto spent fuel storage building. For this approach to work, strong national contribution is needed to facilitate the early communication and exchange of information between the IAEA and the other stakeholders to enable the design of facilities that can be efficiently safeguarded. With the renaissance of nuclear

An examination on aseismatic design of civil engineering structures in nuclear power stations

International Nuclear Information System (INIS)

Aida, Masakazu; Nakamura, Haruaki; Suzuki, Hideya

1990-01-01

As for the aseismatic design of civil engineering structures in nuclear power stations, the basic way of thinking and the example design are shown in the technical guidelines and others, but when the aseismatic design is actually carried out, the techniques of aseismatic calculation and the modeling for structural analysis are left to the judgement of designers. Among such various problems, in this report, the applicability of response displacement method and response magnitude method as the aseismatic calculation techniques and the necessity of rigid region in rahmen members in structural analysis were examined, and those are described. As the structures in nuclear power stations, there are cooling water intake and discharge facilities, emergency water intake, breakwater, unloading quays, revetments, the foundations of tanks, machinery and equipment, piping and others, roads, bridges, tunnels and so on. The outline of their aseismatic design and the examination on the cases of aseismatic design are reported. These structures are frequently underground structures, which are different from those on the ground. (K.I.)

Virtual environments for nuclear power plant design

International Nuclear Information System (INIS)

Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W.

1996-01-01

In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP)

Design optimization for fuel reloading in Laguna Verde nuclear power plant

International Nuclear Information System (INIS)

Cortes Campos, C.C.; Montes Tadeo, J.L.

1991-01-01

Procedure followed to perform the design optimation in fuel reloading is described in general words and also is shown an example in which such procedure was uses for analysis of BWR type reactor in unit 1 of Laguna Verde nuclear power plant (Author)

Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

International Nuclear Information System (INIS)

Hayakawa, Masahiko

1981-01-01

The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Design management and stress analysis of a circular rock tunnel and emplacement holes for storage of spent nuclear fuel

International Nuclear Information System (INIS)

Kandalaft-Ladkany, N.; Wyman, R.V.

1992-01-01

This paper discusses a critical path method (CPM) diagram and logic net which are used for the design cycle of the rock tunnel system for a high level nuclear waste repository. In the analysis the design tunnel is subjected to pre-existing temperature and overburden loads at time of construction. high thermal stresses develop later due to the long term influx of heat from the canisters stored in vertical emplacement holes. Results indicate that thermal stresses reach a critical level for the rock in the vicinity of the canisters which could lead to local collapse of the rock and damage to the canisters

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1983-09-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 3 tables

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1984-02-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 4 tables

Nuclear forensic analysis

International Nuclear Information System (INIS)

Tomar, B.S.

2016-01-01

In the present talk the fundamentals of the nuclear forensic investigations will be discussed followed by the detailed standard operating procedure (SOP) for the nuclear forensic analysis. The characteristics, such as, dimensions, particle size, elemental and isotopic composition help the nuclear forensic analyst in source attribution of the interdicted material, as the specifications of the nuclear materials used by different countries are different. The analysis of elemental composition could be done by SEM-EDS, XRF, CHNS analyser, etc. depending upon the type of the material. Often the trace constituents (analysed by ICP-AES, ICP-MS, AAS, etc) provide valuable information about the processes followed during the production of the material. Likewise the isotopic composition determined by thermal ionization mass spectrometry provides useful information about the enrichment of the nuclear fuel and hence its intended use

Civil design aspects for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Bhalerao, Sandip; Subramanyam, P.; Sharma, Sudin; Bhargava, Kapilesh; Agarwal, Kailash; Rao, D.A.S.; Roy, Amitava; Basu, S.

2015-01-01

The civil design requirements of safety related nuclear structures are much more stringent and conservative as compared to that for conventional and industrial structures. Due to the importance of safety and desired reliability in the civil design of nuclear structures, International Atomic Energy Agency (IAEA) and Atomic Energy Regulatory Board (AERB) have provided various safety guides for their safe design. There has been advancement in theoretical and experimental knowledge pertaining to the design, construction, installation, maintenance, testing and inspection of structures, systems, and components (SSCs) of nuclear power plants (NPPs), such that, their quality and reliability is commensurate with safety functions. The well established procedures are available in the form of different codes, standards, guidelines and well proven research work for NPPs. However, such procedures are somewhat limited in nature for design of civil structures in nuclear fuel cycle facilities (NFCF), and till date no separate codes or standards have been published by regulatory authorities in India that cover civil design aspects for NFCF. Hence, design of civil structures of NFCF in India is performed by using different national and international standards, and the recommendations provided by BARC Safety Council (BSC). Present paper focuses civil design aspects for NFCF in India. (author)

Nuclear safety cooperation for Soviet designed reactors

International Nuclear Information System (INIS)

Reisman, A.W.; Horak, W.C.

1995-01-01

The nuclear accident at the Chernobyl nuclear power plant in 1986 first alerted the West to the significant safety risks of Soviet designed reactors. Five years later, this concern was reaffirmed when the IAEA, as a result of a review by an international team of nuclear safety experts, announced that it did not believe the Kozloduy nuclear power plants in Bulgaria could be operated safely. To address these safety concerns, the G-7 summit in Munich in July 1992 outlined a five point program to address the safety problems of Soviet Designed Reactors: operational safety improvement; near-term technical improvements to plants based on safety assessment; enhancing regulatory regimes; examination of the scope for replacing less safe plants by the development of alternative energy sources and the more efficient use of energy; and upgrading of the plants of more recent design. As of early 1994, over 20 countries and international organizations have pledged hundreds of millions of dollars in financial assistance to improve safety. This paper summarizes these assistance efforts for Soviet designed reactors, draws lessons learned from these activities, and offers some options for better addressing these concerns

Safeguards-by-Design: Early Integration of Physical Protection and Safeguardability into Design of Nuclear Facilities

Energy Technology Data Exchange (ETDEWEB)

T. Bjornard; R. Bean; S. DeMuth; P. Durst; M. Ehinger; M. Golay; D. Hebditch; J. Hockert; J. Morgan

2009-09-01

The application of a Safeguards-by-Design (SBD) process for new nuclear facilities has the potential to minimize proliferation and security risks as the use of nuclear energy expands worldwide. This paper defines a generic SBD process and its incorporation from early design phases into existing design / construction processes and develops a framework that can guide its institutionalization. SBD could be a basis for a new international norm and standard process for nuclear facility design. This work is part of the U.S. DOE’s Next Generation Safeguards Initiative (NGSI), and is jointly sponsored by the Offices of Non-proliferation and Nuclear Energy.

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

International Nuclear Information System (INIS)

T.L. Mitchell

2000-01-01

The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS MandO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS MandO 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS MandO 2000a)

Tornado-resistance design for the nuclear safety structure of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Xia Zufeng.

1987-01-01

The primary design consideration of anti-tornado of the nuclear safety structure of Qinshan Nuclear Power Plant is briefly presented. It mainly includes estimating the probability of tornado arising in the site, ascertaining the design requirments of the anti-tornado structures and deciding the tornado load acted on the structures

Innovative designs for low-level nuclear waste disposal trenches

International Nuclear Information System (INIS)

Nowatzki, E.A.; Armstrong, G.; McCray, J.

1985-01-01

Shallow land burial of low-level nuclear wastes presents many problems that are within the scope of civil engineering analysis and design. These include groundwater seepage, surface water runoff and collection, and the subsidence of trench backfills. Unfortunately, at the time the first disposal sites were being developed, major emphasis was placed on the health-physics aspects of the problem with the result that many of the civil engineering aspects were overlooked and severe problems relating to site integrity exist today. This paper presents the results of a U.S. Nuclear Regulatory Commission (USNRC) sponsored research project conducted at the University of Arizona, Tucson, Arizona, to assess trench cap design from the viewpoint of stability, water infiltration, and economy. Full-scale trenches were constructed that incorporated four different designs. These designs range from a relatively simple cap consisting of engineered backfill with a sloping, compacted soil crown to a more complex cap-crown system that incorporates compacted backfill and a steel reinforced soil-cement cap with an overlaying ''wick'' drain. The results of structural and hydrological monitoring over a period of approximately 15 months are presented. Recommendations are made regarding standard design criteria for future sites based on the results of this research

Nuclear power station: validation of a method for designing operation mimic diagrams

International Nuclear Information System (INIS)

Colard, M.I.; De Vlaminck, M.; Javaux, D.

1995-01-01

This is the first time in a Belgium nuclear power station that the design of mimic diagrams has involved biotechnologists. The methodology is based upon task analysis, the formalism of which has been evaluated. It is now integrated in the Dimos assisted supervision system to produce the final design of mimic diagram images. (authors). 3 figs., 3 refs

Evaluation of earthquake resistance design for underground structures of nuclear power plant, (1)

International Nuclear Information System (INIS)

Tohma, Junichi; Kokusho, Kenji; Iwatate, Takahiro; Ohtomo, Keizo

1986-01-01

As to earthquake resistant design of underground civil engineering structures related with emergency cooling water system of nuclear power plant, it is required these structures must maintain the function of great important their own facilities during earthquakes, especially for design earthquake motion. In this study, shaft pipline, pit and duct for cooling sea water facilities were chosen as typical underground structures, and the authors deal with the seismic design method for calculation of the principal sectional force in these structures generated by design earthquake motion. Especially, comparative investigations concerned with response displacement method versus dynamic analysis methods (lumped mass analysis and finite element analysis) are discussed. (author)

Contributions of the ORNL piping program to nuclear piping design codes and standards

International Nuclear Information System (INIS)

Moore, S.E.

1975-11-01

The ORNL Piping Program was conceived and established to develop basic information on the structural behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design analysis and codes and standards. One of the objectives was to develop and qualify stress indices and flexibility factors for direct use in Code-prescribed design analysis methods. Progress in this area is described

Nuclear fuel elements design, fabrication and performance

CERN Document Server

Frost, Brian R T

1982-01-01

Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

Designing the Nuclear Energy Attitude Scale.

Science.gov (United States)

Calhoun, Lawrence; And Others

1988-01-01

Presents a refined method for designing a valid and reliable Likert-type scale to test attitudes toward the generation of electricity from nuclear energy. Discusses various tests of validity that were used on the nuclear energy scale. Reports results of administration and concludes that the test is both reliable and valid. (CW)

Design management of general contractor under nuclear power project EPC mode

International Nuclear Information System (INIS)

Su Shaojian

2013-01-01

Design management has not yet formed a theoretical system recognized, general contractor design managers under nuclear power project EPC Mode lack the clear theory basis. This paper aims to discuss Design management from the angle of general contractor under nuclear power project EPC mode, Gives the concept of design management Clearly, by Combining the characteristics of nuclear power project, Gives the specific content and meaning of the design management of nuclear power project. (authors)

Multivariable control in nuclear power stations -survey of design methods

International Nuclear Information System (INIS)

Mcmorran, P.D.

1979-12-01

The development of larger nuclear generating stations increases the importance of dynamic interaction between controllers, because each control action may affect several plant outputs. Multivariable control provides the techniques to design controllers which perform well under these conditions. This report is a foundation for further work on the application of multivariable control in AECL. It covers the requirements of control and the fundamental mathematics used, then reviews the most important linear methods, based on both state-space and frequency-response concepts. State-space methods are derived from analysis of the system differential equations, while frequency-response methods use the input-output transfer function. State-space methods covered include linear-quadratic optimal control, pole shifting, and the theory of state observers and estimators. Frequency-response methods include the inverse Nyquist array method, and classical non-interactive techniques. Transfer-function methods are particularly emphasized since they can incorporate ill-defined design criteria. The underlying concepts, and the application strengths and weaknesses of each design method are presented. A review of significant applications is also given. It is concluded that the inverse Nyquist array method, a frequency-response technique based on inverse transfer-function matrices, is preferred for the design of multivariable controllers for nuclear power plants. This method may be supplemented by information obtained from a modal analysis of the plant model. (auth)

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

An intelligent design methodology for nuclear power systems

International Nuclear Information System (INIS)

Nassersharif, B.; Martin, R.P.; Portal, M.G.; Gaeta, M.J.

1989-01-01

The goal of this investigation is to research possible methodologies into automating the design of, specifically, nuclear power facilities; however, it is relevant to all thermal power systems. The strategy of this research has been to concentrate on individual areas of the thermal design process, investigate procedures performed, develop methodology to emulate that behavior, and prototype it in the form of a computer program. The design process has been generalized as follows: problem definition, design definition, component selection procedure, optimization and engineering analysis, testing and final design with the problem definition defining constraints that will be applied to the selection procedure as well as design definition. The result of this research is a prototype computer program applying an original procedure for the selection of the best set of real components that would be used in constructing a system with desired performance characteristics. The mathematical model used for the selection procedure is possibility theory

Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

1998-10-01

Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

Design study of ship based nuclear power reactor

International Nuclear Information System (INIS)

Su'ud, Zaki; Fitriyani, Dian

2002-01-01

Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

Design and analysis of nuclear battery driven by the external neutron source

International Nuclear Information System (INIS)

Wang, Sanbing; He, Chaohui

2014-01-01

Highlights: • A new type of space nuclear power called NBDEx is investigated. • NBDEx with 252 Cf has better performance than RTG with similar structure. • Its thermal power gets great improvement with increment of fuel enrichment. • The service life of NBDEx is about 2.96 year. • The launch abortion accident analysis fully demonstrates the advantage of NBDEx. - Abstract: Based on the theory of ADS (Accelerator Driven Subcritical reactor), a new type of nuclear battery was investigated, which was composed of a subcritical fission module and an isotope neutron source, called NBDEx (Nuclear Battery Driven by External neutron source). According to the structure of GPHS-RTG (General Purpose Heat Source Radioisotope Thermoelectric Generator), the fuel cell model and fuel assembly model of NBDEx were set up, and then their performances were analyzed with MCNP code. From these results, it was found that the power and power density of NBDEx were almost six times higher than the RTG’s. For fully demonstrating the advantage of NBDEx, the analysis of its impact factors was performed with MCNP code, and its lifetime was also calculated using the Origen code. These results verified that NBDEx was more suitable for the space missions than RTG

Overview of seismic resistant design of Indian Nuclear Power Plants

International Nuclear Information System (INIS)

Sharma, G.K.; Hawaldar, R.V.K.P.; Vinod Kumar

2007-01-01

Safe operation of a Nuclear Power Plant (NPP) is of utmost importance. NPPs consist of various Structure, System and Equipment (SS and E) that are designed to resist the forces generated due to a natural phenomenon like earthquake. An earthquake causes severe oscillatory ground motion of short duration. Seismic resistant design of SS and E calls for evaluation of effect of severe ground shaking for assuring the structural integrity and operability during and after the occurrence of earthquake event. Overall exercise is a multi-disciplinary approach. First of standardized 220 MWe design reactor is Narora Atomic Power Station. Seismic design was carried out as per state of art then, for the first time. The twelve 220 MWe reactors and two 540 MWe reactors designed since 1975 have been seismically qualified for the earthquake loads expected in the region. Seismic design of 700 MWe reactor is under advanced stage of finalization. Seismic re-evaluation of six numbers of old plants has been completed as per latest state of art. Over the years, expertise have been developed at Nuclear Power Corporation of India Limited, Bhabha Atomic Research Centre, prominent educational institutes, research laboratories and engineering consultants in the country in the area of seismic design, analysis and shake table testing. (author)

Review of nuclear piping seismic design requirements

International Nuclear Information System (INIS)

Slagis, G.C.; Moore, S.E.

1994-01-01

Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs

Seismic design spectra for nuclear power plants, state-of-the-art

International Nuclear Information System (INIS)

Michalopoulos, A.P.; Shukla, D.K.

1976-01-01

The State-of-the-Art of nuclear power plant design involves the use of design response spectra together with a modal analysis of a mathematical idealization of the actual structure. The design response spectra give the maximum response to ground shaking for a family of single degree-of-freedom viscously damped oscillators. These spectra are usually described as an accelerogram giving ground acceleration as a function of time. The definition of a 'standard' design response spectra is reviewed and illustrated by data relevant to 'hard' or rock sites. Finally, the paper recommends a set of design response spectra applicable to rock sites

Advances in the development of the Mexican platform for analysis and design of nuclear reactors: AZTLAN Platform

International Nuclear Information System (INIS)

Gomez T, A. M.; Puente E, F.; Del Valle G, E.; Francois L, J. L.; Espinosa P, G.

2017-09-01

The AZTLAN platform project: development of a Mexican platform for the analysis and design of nuclear reactors, financed by the SENER-CONACYT Energy Sustain ability Fund, was approved in early 2014 and formally began at the end of that year. It is a national project led by the Instituto Nacional de Investigaciones Nucleares (ININ) and with the collaboration of Instituto Politecnico Nacional (IPN), the Universidad Autonoma Metropolitana (UAM) and Universidad Nacional Autonoma de Mexico (UNAM) as part of the development team and with the participation of the Laguna Verde Nuclear Power Plant, the National Commission of Nuclear Safety and Safeguards, the Ministry of Energy and the Karlsruhe Institute of Technology (Kit, Germany) as part of the user group. The general objective of the project is to modernize, improve and integrate the neutronic, thermo-hydraulic and thermo-mechanical codes, developed in Mexican institutions, in an integrated platform, developed and maintained by Mexican experts for the benefit of Mexican institutions. Two years into the process, important steps have been taken that have consolidated the platform. The main results of these first two years have been presented in different national and international forums. In this congress, some of the most recent results that have been implemented in the platform codes are shown in more detail. The current status of the platform from a more executive view point is summarized in this paper. (Author)

Reactor design for nuclear electric propulsion

International Nuclear Information System (INIS)

Koenig, D.R.; Ranken, W.A.

1979-01-01

Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

2016-08-15

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

Severe accident analysis and management in nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Golshan, Mina

2013-01-01

Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

Seismic design standardization of nuclear facilities

International Nuclear Information System (INIS)

Reddy, G.R.; Vaze, K.K.

2011-01-01

Full text: Structures, Systems and Components (SSCs) of Nuclear Facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Man made accidents such as aircraft impact, explosions etc., some times may be considered as design basis event and some times taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event. It is generally felt design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to be adopted for seismic design standardization of nuclear facilities

Concept of object-oriented intelligent support for nuclear reactor designing

International Nuclear Information System (INIS)

Yoshikawa, H.; Gofuku, A.

1991-01-01

A concept of object-oriented intelligent CAD/CAE environment is proposed for the conceptual designing of advanced nuclear reactor system. It is composed of (i) object-oriented frame-structure database which represents the hierarchical relationship of the composite elements of reactor core and the physical properties, and (ii) object-oriented modularization of the elementary calculation processes, which are needed for reactor core design analysis. As an example practise, an object-oriented frame structure is constructed for representing a 3D configuration of a special fuel element of a space reactor design, by using a general-purpose expert system shell ESHELL/X. (author)

Future needs for inelastic analysis in design of high-temperature nuclear plant components

International Nuclear Information System (INIS)

Corum, J.M.

1980-01-01

The role that inelastic analyses play in the design of high-temperature nuclear plant components is described. The design methodology, which explicitly accounts for nonlinear material deformation and time-dependent failure modes, requires a significant level of realism in the prediction of structural response. Thus, material deformation and failure modeling are, along with computational procedures, key parts of the methodology. Each of these is briefly discussed along with validation by comparisons with benchmark structural tests, and problem areas and needs are discussed for each

Seismic design considerations for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

2001-01-01

During the last few decades, there have been considerable advances in the field of a seismic design of nuclear structures and components housed inside a Nuclear power Plant (NPP). The seismic design and qualification of theses systems and components are carried out through the use of well proven and established theoretical as well as experimental means. Many of the related research works pertaining to these methods are available in the published literature, codes, guides etc. Contrary to this, there is very little information available with regards to the seismic design aspects of the nuclear fuel cycle facilities. This is probably on account of the little importance attached to these facilities from the point of view of seismic loading. In reality, some of these facilities handle a large inventory of radioactive materials and, therefore, these facilities must survive during a seismic event without giving rise to any sort of undue radiological risk to the plant personnel and the public at large. Presented herein in this paper are the seismic design considerations which are adopted for the design of nuclear fuel cycle facilities in India. (author)

Conceptual design of a digital control system for nuclear criticality experiments

International Nuclear Information System (INIS)

Rojas, S.P.

1994-04-01

Nuclear criticality is a concern in many areas of nuclear engineering including waste management, nuclear weapons testing and design, basic nuclear research, and nuclear reactor design and analysis. As in many areas of science and engineering, experimental work conducted in this field has provided a wealth of data and insight essential to the formulation of theory and the advancement in knowledge of fissioning systems. In light of the many diverse applications of nuclear criticality, there is a continuing interest to learn and understand more about the fundamental physical processes through continued experimentation. This thesis addresses the problem of setting up and programming a microprocessor-based digital control system (PLC) for a proposed critical experiment using, among other devices, a stepper motor, a joystick control mechanism, and switches. This experiment represents a revised configuration to test cylindrical nuclear waste packages. A Monte Carlo numerical study for the proposed critical assembly has been performed in order to illustrate how results from numerical calculations are used in the process of assembling the control system and to corroborate previous experimental data. In summary, a control system utilizing some common devices necessary to perform a critical experiment (stepper motor, push-buttons, etc.) has been assembled. Control components were sized using the results of a probabilistic computer code (MCNP). Finally, a program was written that illustrates the coupling between the hardware and the devices being controlled in the new test fixture

Biological analysis with a nuclear microprobe

International Nuclear Information System (INIS)

Cookson, J.A.; Legge, G.J.F.

1975-01-01

Most low-energy nuclear accelerators are now partly used on analytical studies in support of sciences other than nuclear physics. This paper gives a short review of such analytical techniques (X-ray analysis, elastic scattering analysis, nuclear reaction analysis, and the nuclear microprobe) with particular reference to biological applications and also emphasizes the role of the positional analysis that can be performed with a focused beam of ions - the nuclear microprobe. (author)

Genetic algorithms applied to nuclear reactor design optimization

International Nuclear Information System (INIS)

Pereira, C.M.N.A.; Schirru, R.; Martinez, A.S.

2000-01-01

A genetic algorithm is a powerful search technique that simulates natural evolution in order to fit a population of computational structures to the solution of an optimization problem. This technique presents several advantages over classical ones such as linear programming based techniques, often used in nuclear engineering optimization problems. However, genetic algorithms demand some extra computational cost. Nowadays, due to the fast computers available, the use of genetic algorithms has increased and its practical application has become a reality. In nuclear engineering there are many difficult optimization problems related to nuclear reactor design. Genetic algorithm is a suitable technique to face such kind of problems. This chapter presents applications of genetic algorithms for nuclear reactor core design optimization. A genetic algorithm has been designed to optimize the nuclear reactor cell parameters, such as array pitch, isotopic enrichment, dimensions and cells materials. Some advantages of this genetic algorithm implementation over a classical method based on linear programming are revealed through the application of both techniques to a simple optimization problem. In order to emphasize the suitability of genetic algorithms for design optimization, the technique was successfully applied to a more complex problem, where the classical method is not suitable. Results and comments about the applications are also presented. (orig.)

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

International Nuclear Information System (INIS)

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

Institutionalizing Safeguards By Design for Nuclear Facilities

International Nuclear Information System (INIS)

Morgan, James B.; Kovacic, Donald N.; Whitaker, J. Michael

2008-01-01

Safeguards for nuclear facilities can be significantly improved by developing and implementing methodologies for integrating proliferation resistance into the design of new facilities. This paper proposes a method to systematically analyze a facility's processes, systems, equipment, structures and management controls to ensure that all relevant proliferation scenarios that could potentially result in unacceptable consequences have been identified, evaluated and mitigated. This approach could be institutionalized into a country's regulatory structure similar to the way facilities are licensed to operate safely and are monitored through inspections and incident reporting to ensure compliance with domestic and international safeguards. Furthermore, taking credit for existing systems and equipment that have been analyzed and approved to assure a facility's reliable and safe operations will reduce the overall cost of implementing intrinsic and extrinsic proliferation-resistant features. The ultimate goal is to integrate safety, reliability, security and safeguards operations into the design of new facilities to effectively and efficiently prevent diversion, theft and misuse of nuclear material and sensitive technologies at both the facility and state level. To facilitate this approach at the facility level, this paper discusses an integrated proliferation resistance analysis (IPRA) process. If effectively implemented, this integrated approach will also facilitate the application of International Atomic Energy Agency (IAEA) safeguards

Materials for Nuclear Plants From Safe Design to Residual Life Assessments

CERN Document Server

Hoffelner, Wolfgang

2013-01-01

The clamor for non-carbon dioxide emitting energy production has directly  impacted on the development of nuclear energy. As new nuclear plants are built, plans and designs are continually being developed to manage the range of challenging requirement and problems that nuclear plants face especially when managing the greatly increased operating temperatures, irradiation doses and extended design life spans. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments  provides a comprehensive treatment of the structural materials for nuclear power plants with emphasis on advanced design concepts.   Materials for Nuclear Plants: From Safe Design to Residual Life Assessments approaches structural materials with a systemic approach. Important components and materials currently in use as well as those which can be considered in future designs are detailed, whilst the damage mechanisms responsible for plant ageing are discussed and explained. Methodologies for materials characterization, material...

Safety principles and design criteria for nuclear power stations

International Nuclear Information System (INIS)

Gazit, M.

1982-01-01

The criteria and safety principles for the design of nuclear power stations are presented from the viewpoint of a nuclear engineer. The design, construction and operation of nuclear power stations should be carried out according to these criteria and safety principles to ensure, to a reasonable degree, that the likelihood of release of radioactivity as a result of component failure or human error should be minimized. (author)

Preliminary design analysis of hot gas ducts and a intermediate heat exchanger for the nuclear hydrogen reactor

International Nuclear Information System (INIS)

Song, K. N.; Kim, Y. W.

2008-01-01

Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950 .deg. C. Primary and secondary hot gas ducts with coaxial double tubes and are key components connecting a reactor pressure vessel and a intermediate heat exchanger for the nuclear hydrogen system. In this study, preliminary design analyses on the hot gas ducts and the intermediate heat exchanger were carried out. These preliminary design activities include a preliminary design on the geometric dimensions, a preliminary strength evaluation, thermal sizing, and an appropriate material selection

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.; PIEPHO, M.G.

2000-01-01

This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Shield design development of nuclear propulsion merchant ship

International Nuclear Information System (INIS)

Tanaka, Yoshihisa

1975-01-01

Shielding design both in Japan and abroad for nuclear propulsion merchant ships is explained, with emphasis on the various technological problems having occurred in the shield design for one-body type and separate type LWRs as conceptual design. The following matters are described: the peculiarities of the design as compared with the case of land-based nuclear reactors, problems in the design standards of shielding, the present status and development of the design methods, and the instances of the design; thereby, the trends of shielding design are disclosed. The following matters are pointed out: Importance of the optimum design, of shielding, significance of radiation streaming through large voids, activation of the secondary water in built-in type steam generators, and the need of the guides for shield design. (Mori, K.)

Management of design support for nuclear plant modifications

International Nuclear Information System (INIS)

Doyle, F.W.

1991-01-01

The paper will present an overview of the Ontario Hydro organization and processes for providing design support to the operating nuclear power plants. Examples of design support for Pickering GS will be highlighted. The process is described from identification of projects through the design, procurement, construction, commissioning and in-service phases. The practices for managing engineering deliverables are discussed in the context of how these integrate into the overall change control process. The interaction of Engineering with Operations, Construction, Supply and the regulatory bodies is discussed both for major retro-fit programs and for ongoing design support to the nuclear power plants. Recent experiences during the 1990 Pickering Station Outage and during the Unit 3 fuel channel replacement program are highlighted and an integrated 5 year plan for upgrading the safety related systems for the Pickering Nuclear Power Plant is presented. (author)

Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

Science.gov (United States)

Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

2015-04-01

Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic

Time-domain soil-structure interaction analysis of nuclear facilities

International Nuclear Information System (INIS)

Coleman, Justin L.; Bolisetti, Chandrakanth; Whittaker, Andrew S.

2016-01-01

The Nuclear Regulatory Commission (NRC) regulation 10 CFR Part 50 Appendix S requires consideration of soil-structure interaction (SSI) in nuclear power plant (NPP) analysis and design. Soil-structure interaction analysis for NPPs is routinely carried out using guidance provided in the ASCE Standard 4-98 titled “Seismic Analysis of Safety-Related Nuclear Structures and Commentary”. This Standard, which is currently under revision, provides guidance on linear seismic soil-structure-interaction (SSI) analysis of nuclear facilities using deterministic and probabilistic methods. A new appendix has been added to the forthcoming edition of ASCE Standard 4 to provide guidance for time-domain, nonlinear SSI (NLSSI) analysis. Nonlinear SSI analysis will be needed to simulate material nonlinearity in soil and/or structure, static and dynamic soil pressure effects on deeply embedded structures, local soil failure at the foundation-soil interface, nonlinear coupling of soil and pore fluid, uplift or sliding of the foundation, nonlinear effects of gaps between the surrounding soil and the embedded structure and seismic isolation systems, none of which can be addressed explicitly at present. Appendix B of ASCE Standard 4 provides general guidance for NLSSI analysis but will not provide a methodology for performing the analysis. This paper provides a description of an NLSSI methodology developed for application to nuclear facilities, including NPPs. This methodology is described as series of sequential steps to produce reasonable results using any time-domain numerical code. These steps require some numerical capabilities, such as nonlinear soil constitutive models, which are also described in the paper.

Human factors analysis and design methods for nuclear waste retrieval systems. Human factors design methodology and integration plan

Energy Technology Data Exchange (ETDEWEB)

Casey, S.M.

1980-06-01

The purpose of this document is to provide an overview of the recommended activities and methods to be employed by a team of human factors engineers during the development of a nuclear waste retrieval system. This system, as it is presently conceptualized, is intended to be used for the removal of storage canisters (each canister containing a spent fuel rod assembly) located in an underground salt bed depository. This document, and the others in this series, have been developed for the purpose of implementing human factors engineering principles during the design and construction of the retrieval system facilities and equipment. The methodology presented has been structured around a basic systems development effort involving preliminary development, equipment development, personnel subsystem development, and operational test and evaluation. Within each of these phases, the recommended activities of the human engineering team have been stated, along with descriptions of the human factors engineering design techniques applicable to the specific design issues. Explicit examples of how the techniques might be used in the analysis of human tasks and equipment required in the removal of spent fuel canisters have been provided. Only those techniques having possible relevance to the design of the waste retrieval system have been reviewed. This document is intended to provide the framework for integrating human engineering with the rest of the system development effort. The activities and methodologies reviewed in this document have been discussed in the general order in which they will occur, although the time frame (the total duration of the development program in years and months) in which they should be performed has not been discussed.

Human factors analysis and design methods for nuclear waste retrieval systems. Human factors design methodology and integration plan

International Nuclear Information System (INIS)

Casey, S.M.

1980-06-01

The purpose of this document is to provide an overview of the recommended activities and methods to be employed by a team of human factors engineers during the development of a nuclear waste retrieval system. This system, as it is presently conceptualized, is intended to be used for the removal of storage canisters (each canister containing a spent fuel rod assembly) located in an underground salt bed depository. This document, and the others in this series, have been developed for the purpose of implementing human factors engineering principles during the design and construction of the retrieval system facilities and equipment. The methodology presented has been structured around a basic systems development effort involving preliminary development, equipment development, personnel subsystem development, and operational test and evaluation. Within each of these phases, the recommended activities of the human engineering team have been stated, along with descriptions of the human factors engineering design techniques applicable to the specific design issues. Explicit examples of how the techniques might be used in the analysis of human tasks and equipment required in the removal of spent fuel canisters have been provided. Only those techniques having possible relevance to the design of the waste retrieval system have been reviewed. This document is intended to provide the framework for integrating human engineering with the rest of the system development effort. The activities and methodologies reviewed in this document have been discussed in the general order in which they will occur, although the time frame (the total duration of the development program in years and months) in which they should be performed has not been discussed

Development of computer-aided design and production system for nuclear power plant

International Nuclear Information System (INIS)

Ishii, Masanori

1983-01-01

The technically required matters related to the design and production of nuclear power stations tended to increase from the viewpoint of the safety and reliability, and it is indispensable to cope with such technically required matters skillfully for the rationalization of the design and production and for the construction of highly reliable plants. Ishikawajima Harima Heavy Industries Co., Ltd., has developed the computer-aided design data information and engineering system which performs dialogue type design and drawing, and as the result, the design-production consistent system is developed to do stress analysis, production design, production management and the output of data for numerically controlled machine tools consistently. In this paper, mainly the consistent system in the field of plant design centering around piping and also the computer system for the design of vessels and others are outlined. The features of the design works for nuclear power plants, the rationalization of the design and production management of piping and vessels, and the application of the CAD system to other general equipment and improvement works are reported. This system is the powerful means to meet the requirement of heightening quality and reducing cost. (Kako, I.)

Practical design considerations for nuclear cogeneration installations

International Nuclear Information System (INIS)

Koupal, J.R.

1987-01-01

Dual-purpose nuclear plants, cogeneration electricity and steam, offer significant economic benefits over comparable electricity generating stations. The design of such a nuclear facility requires the resolution of unique technical challenges. This paper reports on experience gained in the detailed design of such a dual-purpose facility with the steam supplied to a chemical plant for process heating. The following topics are discussed: Siting, Radioactivity of Export Steam, Optimization for Load Combinations, Steam Supply Reliability, Steam Transportation, Water Chemistry, Cost Allocation. (author)

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

General design criteria for diesel-generator sets for nuclear power plants

International Nuclear Information System (INIS)

Rangarao, G.

1975-01-01

The design criteria for diesel-generators for nuclear power plants are examined. Applicable standards, loading, design performance, and characteristics to be considered in the selection of diesel-generator set and its auxiliary system are discussed. Also, engineered safety features loads together with loss of power safe shutdown loads and their starting sequence, analysis of voltage and frequency response and the diesel-generator ability to start various load blocks successfully to meet the reactor emergency core cooling requirements are discussed

State of the art of earthquake engineering in nuclear power plant design

International Nuclear Information System (INIS)

Schildknecht, P.O.

1976-12-01

A brief outline of definitions based on the USNRC, Seismic and Geologic Siting Criteria for Nuclear Power Plants, and on the plate tectonics and earthquake terminology is given. An introduction into plate tectonics and the associated earthquake phenomena is then presented. Ground motion characteristics are described in connection with the selection of design earthquakes. Mathematical methods of dynamic structural analyses are discussed for linear and nonlinear systems. Response analysis techniques for nuclear power plants are explained considering soil-structure interaction effects. (Auth.)

Safety analysis for key design features of KALIMER-600 design concept

International Nuclear Information System (INIS)

Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

2005-03-01

KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

Quality assurance in the design of nuclear power plants

International Nuclear Information System (INIS)

1981-01-01

This Safety Guide provides the requirements and recommendations related to the establishment and implementation of quality assurance for design of items for a nuclear power plant. The requirements of this Guide shall be applied to the extent necessary during all constituent activities of the nuclear power plant project, such as design, manufacture, construction, commissioning and operations. Its requirements and recommendations shall be implemented, as appropriate, by the responsible organization or by its designated representatives: by plant designers, architect-engineers or manufacturers, when involved in performing design activities related to items to be manufactured; by site constructors, when involved in field engineering activities; by plant operators and other organizations, when involved in design activities related to plant modifications or to selection of spare or replacement parts; and by design consultants and other technical organizations, when performing any engineering activity that affects the work of other design organizations during various stages of nuclear power plant projects

Analysis of the MZA/MZB benchmarks with modern nuclear data sets

International Nuclear Information System (INIS)

Rooijen, W.F.G. van

2013-01-01

Highlights: • ERANOS libraries are produced based on four modern nuclear data sets. • The MOZART MZA/MZB benchmarks are analyzed with these li- braries. • Results are generally acceptable in an academic context, but for highly accurate applications data adjustment is required. • Some discrepancies between the calculations and the benchmark results remain and cannot be readily explained. • Successful generation of ECCO libraries and covariance data for ERA- NOS. - Abstract: For fast reactor design and analysis, our laboratory uses, amongst others, the ERANOS code system. Unfortunately, the publicly available version of ERANOS does not have the most recent nuclear data. Therefore, it was decided to implement an integrated processing system to generate cross sections libraries for the ECCO cell code, as well as covariance data. Cross sections are generated from the original ENDF files. For our purposes, it is important to ascertain that the ECCO cross section libraries are of adequate quality to allow design and analysis of advanced fast reactors in an academic context. In this paper, we present an analysis of the MZA/MZB benchmarks with nuclear data from JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. Results are that reactivity is generally well predicted, with an uncertainty of about 1% due to covariances of the nuclear data. Reaction rate ratios are satisfactorily calculated, as well as the flux spectrum and reaction rate traverses. Some problems remain: the magnitude of the void effect is not satisfactorily calculated, and reaction rate traverses are not always satisfactorily calculated. On the whole, the ECCO libraries are sufficient for design and analysis tasks in an academic context. For high-precision calculations, such as required for licensing tasks and detailed design calculations, data adjustment is still necessary as the “native” covariance data in the ENDF files is not accurate enough

Design of particle bed reactors for the space nuclear thermal propulsion program

International Nuclear Information System (INIS)

Ludewig, H.; Powell, J.R.; Todosow, M.; Maise, G.; Barletta, R.; Schweitzer, D.G.

1996-01-01

This paper describes the design for the Particle Bed Reactor (PBR) that was considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated. (author)

Structural design of nuclear reactor machinery and equipment

International Nuclear Information System (INIS)

Hara, Hideki

1992-01-01

Since the machinery, equipment and piping which compose nuclear power station facilities are diverse, when those are designed, consideration is given sufficiently to the objective of use and the importance of the object machinery and equipment so that those can maintain the soundness over the design life. In this report, on the contents and the design standard in the design techniques for nuclear reactor machinery and equipment, the way of thinking is shown, taking an example of reactor pressure vessel which is stipulated as the vessel kind 1 in the 'Technical standard of structures and others regarding nuclear facilities for electric power generation', Notice No. 501 of the Ministry of International Trade and Industry. The reactor pressure vessel of 1350 MWe improved type BWR (ABWR) is used under the condition of 87.9 kg/cm 2 and 302 degC, and the inside diameter is about 7.2 m, the inside height is about 21 m, and the wall thickness is about 170 mm. The design standard for reactor pressure vessels and its way of thinking, breakdown prevention design and the design techniques for reactor pressure vessels are described. (K.I.)

Earthquake resistant design of nuclear facilities with limited radioactive inventory

International Nuclear Information System (INIS)

1985-10-01

This document comprises the essential elements of an earthquake resistant design code for nuclear facilities with limited radioactive inventory. The purpose of the document is the enhancement of seismic safety for such facilities without the necessity to resort to complicated and sophisticated methodologies which are often associated with and borrowed from nuclear power plant analysis and design. The first two sections are concerned with the type of facility for which the document is applicable and the radiological consideration for accident conditions. The principles of facility classification and item categorization as a function of the potential radiological consequences of failure are given in section 3. The design basis ground motion is evaluated in sections 4-6 using a simplified but conservative approach which also includes considerations for the underlying soil characteristics. Sections 7 and 8 specify the principles of seismic design of building structures and equipment using two methods, called the equivalent static and simplified dynamic approach. Considerations for the detailing of equipment and piping and those other than for lateral load calculations, such as sloshing effects, are given in the subsequent sections. Several appendices are given for illustration of the principles presented in the text. Finally, a design tree diagram is included to facilitate the user's task of making the appropriate selections. (author)

Conceptual design of technical security systems for Russian nuclear facilities physical protection

International Nuclear Information System (INIS)

Izmailov, A.V.

1995-01-01

Conceptual design of technical security systems (TSS) used in the early stages of physical protection systems (PPS) design for Russia nuclear facilities is discussed. The importance of work carried out in the early stages was noted since the main design solutions are being made within this period (i.e. selection of a structure of TSS and its components). The methods of analysis and synthesis of TSS developed by ''Eleron'' (MINATOM of Russia) which take into account the specific conditions of Russian nuclear facilities and a scope of equipment available are described in the review. TSS effectiveness assessment is based on a probability theory and a simulation. The design procedure provides for a purposeful choice of TSS competitive options including a ''cost-benefit'' criterion and taking into account a prechosen list of design basis threats to be used for a particular facility. The attention is paid to a practical aspect of the methods application as well as to the bilateral Russian-American scientific and technical co-operation in the PPS design field

Seismic fragility analysis of a nuclear building based on probabilistic seismic hazard assessment and soil-structure interaction analysis

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, R.; Ni, S.; Chen, R.; Han, X.M. [CANDU Energy Inc, Mississauga, Ontario (Canada); Mullin, D. [New Brunswick Power, Point Lepreau, New Brunswick (Canada)

2016-09-15

Seismic fragility analyses are conducted as part of seismic probabilistic safety assessment (SPSA) for nuclear facilities. Probabilistic seismic hazard assessment (PSHA) has been undertaken for a nuclear power plant in eastern Canada. Uniform Hazard Spectra (UHS), obtained from the PSHA, is characterized by high frequency content which differs from the original plant design basis earthquake spectral shape. Seismic fragility calculations for the service building of a CANDU 6 nuclear power plant suggests that the high frequency effects of the UHS can be mitigated through site response analysis with site specific geological conditions and state-of-the-art soil-structure interaction analysis. In this paper, it is shown that by performing a detailed seismic analysis using the latest technology, the conservatism embedded in the original seismic design can be quantified and the seismic capacity of the building in terms of High Confidence of Low Probability of Failure (HCLPF) can be improved. (author)

Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

International Nuclear Information System (INIS)

Elder, J.; Graf, J.M.

1984-01-01

DOE Office of Nuclear Safety has sponsored preparation of a guidance document to aid field offices and contractors in their analyses of consequences of postulated major accidents. The guide addresses the requirements of DOE Orders 5480.1A, Chapter V, and 6430.1, including the general requirement that DOE nuclear facilities be sited, designed, and operated in accordance with standards, codes, and guides consistent with those applied to comparable licensed nuclear facilities. The guide includes both philosophical and technical information in the areas of: siting guidelines doses applied to an offsite reference person; consideration also given to an onsite reference person; physical parameters, models, and assumptions to be applied when calculating doses for comparison to siting criteria; and potential accident consequences other than radiological dose to a reference person which might affect siting and major design features of the facility, such as environmental contamination, population dose, and associated public health effects. Recommendations and/or clarifications are provided where this could be done without adding new requirements. In this regard, the guide is considered a valuable aid to the safety analyst, especially where requirements have been subject to inconsistent interpretation or where analysis methods are in transition, such as use of dose model (ICRP 2 or ICRP 30) or use of probabilistic methods of risk analysis in the siting and design of nuclear facilities

Nuclear challenges and progress in designing stellarator fusion power plants

International Nuclear Information System (INIS)

El-Guebaly, L.A.; Wilson, P.; Henderson, D.; Sawan, M.; Sviatoslavsky, G.; Tautges, T.; Slaybaugh, R.; Kiedrowski, B.; Ibrahim, A.

2008-01-01

Over the past 5-6 decades, stellarator power plants have been studied in the US, Europe, and Japan as an alternate to the mainline magnetic fusion tokamaks, offering steady-state operation and eliminating the risk of plasma disruptions. The earlier 1980s studies suggested large-scale stellarator power plants with an average major radius exceeding 20 m. The most recent development of the compact stellarator concept delivered ARIES-CS - a compact stellarator with 7.75 m average major radius, approaching that of tokamaks. For stellarators, the most important engineering parameter that determines the machine size and cost is the minimum distance between the plasma boundary and mid-coil. Accommodating the breeding blanket and necessary shield within this distance to protect the ARIES-CS superconducting magnet represents a challenging task. Selecting the ARIES-CS nuclear and engineering parameters to produce an economic optimum, modeling the complex geometry for 3D nuclear analysis to confirm the key parameters, and minimizing the radwaste stream received considerable attention during the design process. These engineering design elements combined with advanced physics helped enable the compact stellarator to be a viable concept. This paper provides a brief historical overview of the progress in designing stellarator power plants and a perspective on the successful integration of the nuclear activity into the final ARIES-CS configuration

Design of a quality assurance system in the nuclear fuel fabrication

International Nuclear Information System (INIS)

Garcia Rojas Palacios, L.

1992-01-01

A)For the first time a project on nuclear fuel fabrication is going to be lead in this country. For this reason the work is oriented to establish a quality assurance system for the different stages of fuel fabrication. C) The work of this thesis was developed first by means of an analysis of quality philosophies of Deming, Ishikawa, Juran and Crosby from which several important points were stracted to be used in the designed quality system. Metrology and normalization are so important for quality control that a study of them is made considering definitions, unit systems and type of errors (for Metrology) as well as standards for quality systems, qualification, destructive and non destructive tests, shipment, packing for nuclear power plants. With the standards as a basis, the working strategy for the system was reached, as well as the design of control cards and the design of documents for inspection control, personnel and its documentation and finally the diagrams for each one of the fabrication stages

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

Development of an event-driven parser for active document and web-based nuclear design system

Energy Technology Data Exchange (ETDEWEB)

Park, Yong Soo

2005-02-15

Nuclear design works consist of extensive unit job modules in which many computer codes are used. Each unit module requires time-consuming and erroneous input preparation, code run, output analysis and quality assurance process. The task for safety evaluation of reload core is especially the most man-power intensive and time-consuming due to the large amount of calculations and data exchanges. The purpose of this study is to develop a new nuclear design system called Innovative Design Processor (IDP) in order to minimize human effort and maximize design quality and productivity, and then to achieve an ultimately optimized core loading pattern. Two new basic principles of IDP are the document-oriented design and the web based design. Contrary to the conventional code-oriented or procedure-oriented design, the document-oriented design is human-oriented in that the final document is automatically prepared with complete analysis, table and plots, if the designer writes a design document called active document and feeds it to a parser. This study defined a number of active components and developed an event-driven parser for the active document in HTML (Hypertext Markup Language) or XML (Extensible Markup Language). The active documents can be created on the web, which is another framework of IDP. Using proper mix-up of server side and client side programming under the HAMP (HP-UX/Apache/MySQL/PHP) environment, the document-oriented design process on the web is modeled as a design wizard for designer's convenience and platform independency. This automation using IDP was tested for the reload safety evaluation of Korea Standard Nuclear Power Plant (KSNP) type PWRs. Great time saving was confirmed and IDP can complete several-month jobs in a few days. More optimized core loading pattern, therefore, can be obtained since it takes little time to do the reload safety evaluation tasks with several core loading pattern candidates. Since the technology is also applicable to

Development of an event-driven parser for active document and web-based nuclear design system

International Nuclear Information System (INIS)

Park, Yong Soo

2005-02-01

Nuclear design works consist of extensive unit job modules in which many computer codes are used. Each unit module requires time-consuming and erroneous input preparation, code run, output analysis and quality assurance process. The task for safety evaluation of reload core is especially the most man-power intensive and time-consuming due to the large amount of calculations and data exchanges. The purpose of this study is to develop a new nuclear design system called Innovative Design Processor (IDP) in order to minimize human effort and maximize design quality and productivity, and then to achieve an ultimately optimized core loading pattern. Two new basic principles of IDP are the document-oriented design and the web based design. Contrary to the conventional code-oriented or procedure-oriented design, the document-oriented design is human-oriented in that the final document is automatically prepared with complete analysis, table and plots, if the designer writes a design document called active document and feeds it to a parser. This study defined a number of active components and developed an event-driven parser for the active document in HTML (Hypertext Markup Language) or XML (Extensible Markup Language). The active documents can be created on the web, which is another framework of IDP. Using proper mix-up of server side and client side programming under the HAMP (HP-UX/Apache/MySQL/PHP) environment, the document-oriented design process on the web is modeled as a design wizard for designer's convenience and platform independency. This automation using IDP was tested for the reload safety evaluation of Korea Standard Nuclear Power Plant (KSNP) type PWRs. Great time saving was confirmed and IDP can complete several-month jobs in a few days. More optimized core loading pattern, therefore, can be obtained since it takes little time to do the reload safety evaluation tasks with several core loading pattern candidates. Since the technology is also applicable to the

Design and analysis of new prestressed concrete containment and its passive cooling system for nuclear power plants

International Nuclear Information System (INIS)

Tan Xiaoshi; Li Xiaowei; Li Xiaotian; He Shuyan

2014-01-01

A new nuclear power plant prestressed concrete containment and its passive cooling system design were proposed for CAP1700 nuclear power plant as an example. The thermal-hydraulic calculation method for the new passive containment cooling system of CAP1700 was introduced and the operating parameters in accident condition were obtained. The result shows that the design of passive containment cooling system for CAP1700 is feasible and can meet the cooling demand in accident condition. Reservoir capacity of tank has a big margin and can be further optimized by calculation. (authors)

Fundamentals for the safety-relevant design of rotational shiftplans in nuclear power plants

International Nuclear Information System (INIS)

Preuss, W.; Herbert, K.W.; Reinartz, G.; Saniter, R.

1984-01-01

The investigation was concerned with the influence of rotational shiftwork on operator performance, particularly in the nuclear power industry. In an analysis of the literature, the fundamental chronobiological knowledge was reviewed. Work physiological criteria for the design of shiftplans were assembled and their relevance to this industry was assessed. In addition, interviews were held with representatives from nuclear power utilities in the Federal Republic of Germany and also in Switzerland and the Netherlands. Characteristics of the shiftplans in current use were recorded and the basic operating principles of the shift systems were ascertained. The alternative designs were discussed and assessed from the point of view of operator performance and their significance to safety. The results of the analysis of the literature and of the interviews with the utilities are presented in two separate volumes of the report. (orig.) [de

Establishment of a clean laboratory for ultra trace analysis of nuclear materials in safeguards environmental samples

International Nuclear Information System (INIS)

Hanzawa, Yukiko; Magara, Masaaki; Watanabe, Kazuo

2003-01-01

The Japan Atomic Energy Research Institute has established a cleanroom facility with cleanliness of ISO Class 5: the Clean Laboratory for Environmental Analysis and Research (CLEAR). It was designed to be used for the analysis of nuclear materials in environmental samples mainly for the safeguards, in addition to the Comprehensive Nuclear-Test-Ban Treaty verification and research on environmental sciences. The CLEAR facility was designed to meet conflicting requirements of a cleanroom and for handling of nuclear materials according to Japanese regulations, i.e., to avoid contamination from outside and to contain nuclear materials inside the facility. This facility has been intended to be used for wet chemical treatment, instrumental analysis and particle handling. A fume-hood to provide a clean work surface for handling of nuclear materials was specially designed. Much attention was paid to the selection of construction materials for use to corrosive acids. The performance of the cleanroom and analytical background in the laboratory are discussed. This facility has satisfactory specification required for joining the International Atomic Energy Agency Network of Analytical Laboratories. It can be concluded that the CLEAR facility enables analysis of ultra trace amounts of nuclear materials at sub-pictogram level in environmental samples. (author)

New nuclear plant design and licensing process

International Nuclear Information System (INIS)

Luangdilok, W.

1996-01-01

This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

International Nuclear Information System (INIS)

Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

1998-10-01

Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

Nuclear renaissance, public perception and design criteria: An exploratory review

International Nuclear Information System (INIS)

Goodfellow, Martin J.; Williams, Hugo R.; Azapagic, Adisa

2011-01-01

There is currently an international drive to build new nuclear power plants, bringing about what is being termed a 'nuclear renaissance'. However, the public perception of nuclear energy has historically been, and continues to be, a key issue, particularly in light of the Fukushima nuclear incident. This paper discusses the disparity between perceived and calculated risks based on the last four decades of research into risk perception. The leading psychological and sociological theories, Psychometric Paradigm and Cultural Theory, respectively, are critically reviewed. The authors then argue that a new nuclear-build policy that promotes a broader approach to design incorporating a wider range of stakeholder inputs, including that of the lay public, may provide a means for reducing the perceived risk of a nuclear plant. Further research towards such a new approach to design is proposed, based on integrating expert and lay stakeholder inputs and taking into account broader socio-cultural factors whilst maintaining the necessary emphasis on safety, technological development, economics and environmental sustainability. - Highlights: → Globally, a number of countries are investing in or considering building new nuclear plants. → Public acceptance of nuclear safety is important to continuing new nuclear build efforts. → Theories are discussed attempting to explain the public perception of nuclear safety. → A socially informed design process is proposed which could assist in ensuring public support. → Further research to understand how this design process might be performed is proposed.

performance-based approach to design and evaluation of nuclear security systems for Brazilian nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Tavares, Renato L. A.; Filho, Josélio S. M., E-mail: renato.tavares@cnen.gov.br, E-mail: joselio@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Diretoria de Radioproteção e Segurança Nuclear. Divisão de Normas e Segurança Física; Fontes, Gladson S.; Fiel, J.C.B., E-mail: gsfontes@hotmail.com, E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (SE-7/IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

This study presents an application of a performance-based approach to definition of requirements, design and evaluation of physical protection systems for nuclear facilities. Such approach considers a probabilistic analysis of the threat, equipment, systems and response forces used to prevent, dissuade and detain malicious acts against the integrity of facilities and the nuclear materials inside them. Nowadays, in the context of Brazilian nuclear facilities licensing, a mostly prescriptive approach is adopted, which despite having advantages such as simplified inspections and homogeneous regulatory requisites amid different fuel cycle facility types, does not consider evolution, dynamism and capacities of external or internal threats to facilities and to Brazilian Nuclear Program itself, neither provides metrics to evaluate system performance facing such threats. In order to preserve actual plans and systems confidentiality, a facility hypothetical model is created, including a research reactor and a waste storage facility. It is expected that the methodology and results obtained in this study serve in the future as a basis to Brazilian nuclear operators, in elaboration process of their Physical Protection Plans, which must comply with future regulation CNEN-NN 2.01, a revision of CNEN-NE 2.01, once that regulation will include performance requisites. (author)

performance-based approach to design and evaluation of nuclear security systems for Brazilian nuclear facilities

International Nuclear Information System (INIS)

Tavares, Renato L. A.; Filho, Josélio S. M.; Fontes, Gladson S.; Fiel, J.C.B.

2017-01-01

This study presents an application of a performance-based approach to definition of requirements, design and evaluation of physical protection systems for nuclear facilities. Such approach considers a probabilistic analysis of the threat, equipment, systems and response forces used to prevent, dissuade and detain malicious acts against the integrity of facilities and the nuclear materials inside them. Nowadays, in the context of Brazilian nuclear facilities licensing, a mostly prescriptive approach is adopted, which despite having advantages such as simplified inspections and homogeneous regulatory requisites amid different fuel cycle facility types, does not consider evolution, dynamism and capacities of external or internal threats to facilities and to Brazilian Nuclear Program itself, neither provides metrics to evaluate system performance facing such threats. In order to preserve actual plans and systems confidentiality, a facility hypothetical model is created, including a research reactor and a waste storage facility. It is expected that the methodology and results obtained in this study serve in the future as a basis to Brazilian nuclear operators, in elaboration process of their Physical Protection Plans, which must comply with future regulation CNEN-NN 2.01, a revision of CNEN-NE 2.01, once that regulation will include performance requisites. (author)

Nuclear data library in design calculation

International Nuclear Information System (INIS)

Hirano, Go; Kosaka, Shinya

2006-01-01

In core design calculation, nuclear data takes part as multi group cross section library during the assembly calculation, which is the first stage of a core design calculation. This report summarizes the multi group cross section libraries used in assembly calculations and also presents the methods adopted for resonance and assembly calculation. (author)

Nuclear analysis of the ITER Cryopump Ports

International Nuclear Information System (INIS)

Moro, Fabio; Villari, Rosaria; Flammini, Davide; Antipenkov, Alexander; Dremel, Matthias; Levesy, Bruno; Loughlin, Michael; Juarez, Rafael; Perez, Lucia; Petrizzi, Luigino

2015-01-01

Highlights: • Evaluation the shielding effectiveness of the TCPHs by means of 3-D neutrons and gamma maps. • Assessment of the nuclear heating induced by neutron and photons on the TCP and TCPHs. • Calculation of the dose rate at 12 days after shutdown in the maintenance area of the Lower Ports with the Advanced D1S method, in order to verify the design target (100 μSv/h). • Potential improvements of the shielding configuration aimed at the reduction of the dose level in the Port Cell have been proposed and discussed. - Abstract: The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.

Nuclear analysis of the ITER Cryopump Ports

Energy Technology Data Exchange (ETDEWEB)

Moro, Fabio, E-mail: fabio.moro@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Flammini, Davide [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Antipenkov, Alexander; Dremel, Matthias; Levesy, Bruno; Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Juarez, Rafael; Perez, Lucia [UNED, Energetic Engineering Department, C/Juan del Rosal 12, Madrid (Spain); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium)

2015-10-15

Highlights: • Evaluation the shielding effectiveness of the TCPHs by means of 3-D neutrons and gamma maps. • Assessment of the nuclear heating induced by neutron and photons on the TCP and TCPHs. • Calculation of the dose rate at 12 days after shutdown in the maintenance area of the Lower Ports with the Advanced D1S method, in order to verify the design target (100 μSv/h). • Potential improvements of the shielding configuration aimed at the reduction of the dose level in the Port Cell have been proposed and discussed. - Abstract: The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.

Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

International Nuclear Information System (INIS)

Ambrosini, Daniel; Codina, Ramón H.; Curadelli, Oscar; Martínez, Carlos A.

2017-01-01

Highlights: • Structural analysis of CAREM-25 NPP is presented. • Full 3D numerical model was developed. • Transient thermal and static structural analyses were performed. • Modeling guidelines for numerical structural analysis of NPP are recommended. • Envelope condition of DBA dominates the structural behavior. - Abstract: In this paper, a numerical study about the structural response of the Argentine nuclear power plant CAREM-25 subjected to the design basis accident (DBA) and seismic loads is presented. Taking into account the hardware capabilities available, a full 3D finite element model was adopted. A significant part of the building was modeled using more than 2 M solid elements. In order to take into account the foundation flexibility, linear springs were used. The springs and the model were calibrated against a greater model used to study the soil-structure interaction. The structure was subjected to the DBA and seismic loads as combinations defined by ASME international code. First, a transient thermal analysis was performed with the conditions defined by DBA and evaluating the time history of the temperature of the model, each 1 h until 36 h. The final results of this stage were considered as initial conditions of a static structural analysis including the pressure defined by DBA. Finally, an equivalent static analysis was performed to analyze the seismic response considering the design basis spectra for the site. The different loads were combined and the abnormal/extreme environmental combination was the most unfavorable for the structure, defining the design.

Seismic analysis of liquid storage container in nuclear reactors

International Nuclear Information System (INIS)

Zhang Zhengming; He Shuyan; Xu Ming

2007-01-01

Seismic analysis of liquid storage containers is always difficult in the seismic design of nuclear reactor equipment. The main reason is that the liquid will generate significant seismic loads under earthquake. These dynamic liquid loads usually form the main source of the stresses in the container. For this kind of structure-fluid coupling problem, some simplified theoretical methods were usually used previously. But this cannot satisfy the requirements of engineering design. The Finite Element Method, which is now full developed and very useful for the structural analysis, is still not mature for the structure-fluid coupling problem. This paper introduces a method suitable for engineering mechanical analysis. Combining theoretical analysis of the dynamic liquid loads and finite element analysis of the structure together, this method can give practical solutions in the seismic design of liquid storage containers

Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

International Nuclear Information System (INIS)

Elder, J.C.; Graf, J.M.; Dewart, J.M.; Buhl, T.E.; Wenzel, W.J.; Walker, L.J.; Stoker, A.K.

1986-01-01

This guide was prepared to provide the experienced safety analyst with accident analysis guidance in greater detail than is possible in Department of Energy (DOE) Orders. The guide addresses analysis of postulated serious accidents considered in the siting and selection of major design features of DOE nuclear facilities. Its scope has been limited to radiological accidents at nonreactor nuclear facilities. The analysis steps addressed in the guide lead to evaluation of radiological dose to exposed persons for comparison with siting guideline doses. Other possible consequences considered are environmental contamination, population dose, and public health effects. Choices of models and parameters leading to estimation of source terms, release fractions, reduction and removal factors, dispersion and dose factors are discussed. Although requirements for risk analysis have not been established, risk estimates are finding increased use in siting of major nuclear facilities, and are discussed in the guide. 3 figs., 9 tabs

The impact of safety standards updating for design purposes in nuclear power plants licensing

International Nuclear Information System (INIS)

Alvarenga, Marco Antonio Bayout; Rabello, Sidney Luiz

2009-01-01

The Brazilian experience of nuclear power plants licensing was consolidated by the use of the Brazilian, American, German and IAEA standards. Independently of the set of norms, standards or guides to be used, this set should be in consonance with the state-of-art or the current state of knowledge in science and technology. In the general design criteria of US NRC or German BMI, or in the Brazilian norms (CNEN) or even, in the IAEA standards, this aspect is always emphasized. On the other hand, the international operational experience of nuclear reactors (for example, TMI accident) also contributes to the updating of norms and standards. The use of new technologies (for example, digital technology) impels the norms and standards to adopt new design criteria related to the new technological context. Moreover, we must add the particular vision that each country can have concerning to specific topics in nuclear safety. This work discusses how the norms, standards and guides used in the nuclear licensing are being reviewed to cope with the requirement of the state-of-art. In order to accomplish this aim we took some general design criteria to exemplify how they are fulfilled, mainly those related directly with the protection of the defense-in-depth barriers: primary coolant system, containment vessel and containment systems, including external events and severe accidents. In complement to the deterministic analysis, it is also discussed the design criteria related to the human factors engineering and probabilistic safety analysis, including severe accidents aspects. (author)

Reliability analysis of scram system of a critical nuclear power plant

International Nuclear Information System (INIS)

Vieira Neto, A.S.; Souza Borges, W. de

1986-01-01

The object of this paper is to show the relevancy of reliability analysis of nuclear systems as a mean of evaluating their prospect performance in design phase. For this purpose a typical scram system design for light water cooled critical facilities is analized to verify the effects of alternative maintenance procedure and design redundancies in realibility characteristics. (Author) [pt

Nuclear Reactor Engineering Analysis Laboratory

International Nuclear Information System (INIS)

Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

1998-01-01

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

Thermal analysis of NNWSI conceptual waste package designs

International Nuclear Information System (INIS)

Stein, W.; Hockman, J.N.; O'Neal, W.C.

1984-04-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This report discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 16 references

A framework for AI-based nuclear design support system

International Nuclear Information System (INIS)

Furuta, Kazuo; Kondo, Shunsuke

1991-01-01

Nowadays many computer programs are being developed and used for the analytic tasks in nuclear reactor design, but experienced designers are still responsible for most of the synthetic tasks which are not amenable to algorithmic computer processes. Artificial intelligence (AI) is a promising technology to deal with these intractable tasks in design. In development of AI-based design support systems, it is desirable to choose a comprehensive framework based on the scientific theory of design. In this work a framework for AI-based design support systems for nuclear reactor design will be proposed based on an exploration model of design. The fundamental architectures of this framework will be described especially on knowledge representation, context management and design planning. (author)

Status of CONRAD, a nuclear reaction analysis tool

International Nuclear Information System (INIS)

Saint Jean, C. de; Habert, B.; Litaize, O.; Noguere, G.; Suteau, C.

2008-01-01

The development of a software tool (CONRAD) was initiated at CEA/Cadarache to give answers to various problems arising in the data analysis of nuclear reactions. This tool is then characterized by the handling of uncertainties from experimental values to covariance matrices for multi-group cross sections. An object oriented design was chosen allowing an easy interface with graphical tool for input/output data and being a natural framework for innovative nuclear models (Fission). The major achieved developments are a data model for describing channels, nuclear reactions, nuclear models and processes with interface to classical data formats, theoretical calculations for the resolved resonance range (Reich-Moore) and unresolved resonance range (Hauser-Feshbach, Gilbert-Cameron,...) with nuclear model parameters adjustment on experimental data sets and a Monte Carlo method based on conditional probabilities developed to calculate properly covariance matrices. The on-going developments deal with the experimental data description (covariance matrices) and the graphical user interface. (authors)

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Nuclear Design Study for Reduced Boron Concentration Operation

International Nuclear Information System (INIS)

Daing, Aung Tharn; Kim, Myung Hyun

2008-01-01

In PWR, Boron Dilution Accident (BDA) which is an inadvertent reduction of the soluble boron concentration in the primary coolant, can lead to a positive reactivity insertion and a power increase. In this study, an optimal Low Boron Concentration (LBC) core design will be developed by reducing an excess reactivity so as to possess passive inherently safety features as well as one of attentively preventive measures of BDA scenarios. The scope of the research project concerning the change of operational procedure in chemical shim system, will involve three main parts namely, nuclear design for LBC core, feasibility check on the proposed core in safety and costs, and PSA approach on design change benefits. In addition, successful system analysis of the reactivity effects due to boron dilution transients, require a full coupling between core thermal hydraulic, reactor dynamics, and the use of Computational Fluid Dynamics (CFD) codes

Nuclear Design Study for Reduced Boron Concentration Operation

Energy Technology Data Exchange (ETDEWEB)

Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2008-10-15

In PWR, Boron Dilution Accident (BDA) which is an inadvertent reduction of the soluble boron concentration in the primary coolant, can lead to a positive reactivity insertion and a power increase. In this study, an optimal Low Boron Concentration (LBC) core design will be developed by reducing an excess reactivity so as to possess passive inherently safety features as well as one of attentively preventive measures of BDA scenarios. The scope of the research project concerning the change of operational procedure in chemical shim system, will involve three main parts namely, nuclear design for LBC core, feasibility check on the proposed core in safety and costs, and PSA approach on design change benefits. In addition, successful system analysis of the reactivity effects due to boron dilution transients, require a full coupling between core thermal hydraulic, reactor dynamics, and the use of Computational Fluid Dynamics (CFD) codes.

Summary of trial design of improved marine nuclear reactors

International Nuclear Information System (INIS)

1984-01-01

In order to carry out the research and development of improved marine nuclear reactors, the Japan Nuclear Ship Research and Development Agency decided the project for the purpose in accordance with the procedure of research and development shown by the Nuclear Ship Research and Development Committee of Atomic Energy Commission in December, 1979, and along the basic plan regarding the development of nuclear ships of the Agency decided in February, 1981. As the first step, the Agency has been advancing the research on the design evaluation comprising the trial design and conceptual design to establish the concept of the marine reactor plant with excellent economical efficiency and reliability, which will be developed as the practical plant for future nuclear ships. The trial design started as a three-year project from 1983 is related to a 100 MWt marine reactor, and it is to obtain the concept of improved marine reactors which can be realized after adequate development period based on the pressurized water reactors of separate type, one-body type and semi-one-body type. In this summary, the works carried out in fiscal year 1983 are reported, that is, the design and calculation of the reactor core and the equipment of primary cooling system, and the selection of the required items of research and development. (Kako, I.)

Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

International Nuclear Information System (INIS)

Skogen, F.B.; Stout, R.B.

1977-01-01

Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

Human factor engineering applied to nuclear power plant design

International Nuclear Information System (INIS)

Manrique, A.; Valdivia, J.C.; Jimenez, A.

2001-01-01

For the design and construction of new nuclear power plants as well as for maintenance and operation of the existing ones new man-machine interface designs and modifications are been produced. For these new designs Human Factor Engineering must be applied the same as for any other traditional engineering discipline. Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. Additionally, the big saving achieved by a nuclear power plant having an operating methodology which significantly decreases the risk of operating errors makes it necessary and almost vital its implementation. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. (author)

Fast breeder physics and nuclear core design

International Nuclear Information System (INIS)

Marth, W.; Schroeder, R.

1983-07-01

This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

Methodology for modular nuclear plant design and construction

International Nuclear Information System (INIS)

Lapp, C.W.; Golay, M.

1992-01-01

During the past decade, the rising cost of nuclear power plant construction has caused the cancellation of many projects and has forced some utilities into bankruptcy. Many factors have contributed to capital cost increases, including regulatory changes, the absence of standard designs, and low worker productivity. Low worker productivity can be attributed to the conventional building process, which is not conductive to productive labor. This study presents innovative ways to reduce the capital cost of nuclear plants through more efficient construction processes designed to increase worker productivity. A major portion of the plant capital cost is the interest paid during construction on borrowed capital. Modular fabrication could potentially reduce interest payments by compressing the construction schedule of nuclear facilities. Additional cost savings expected from modular designs arise from improved quality, productivity, and schedule control in fabrication of plant elements within a factory environment

Heat balance calculation and feasibility analysis for initial startup of Fuqing nuclear turbine unit with non-nuclear steam

International Nuclear Information System (INIS)

He Liu; Xiao Bo; Song Yumeng

2014-01-01

Non-nuclear steam run up compared with nuclear steam run up, can verify the design, manufacture, installation quality of the unit, at the same time shorten the follow-up duration of the entire group ready to start debugging time. In this paper, starting from the first law of thermodynamics, Analyzed Heat balance Calculation and Feasibility analysis for Initial startup of Fuqing nuclear Turbine unit with Non-nuclear steam, By the above calculation, to the system requirements and device status on the basis of technical specifications, confirmed the feasibility of Non-nuclear steam running up in theory. After the implementation of the Non-nuclear turn of Fuqing unit, confirmed the results fit with the actual process. In summary, the Initial startup of Fuqing turbine unit with Non-nuclear steam is feasible. (authors)

Nuclear power. Volume 2. Nuclear power project management

International Nuclear Information System (INIS)

Pedersen, E.S.

1978-01-01

NUCLEAR POWER PLANT DESIGN is intended to be used as a working reference book for management, engineers and designers, and as a graduate-level text for engineering students. The book is designed to combine theory with practical nuclear power engineering and design experience, and to give the reader an up-to-date view of the status of nuclear power and a basic understanding of how nuclear power plants function. Volume 2 contains the following chapters: (1) review of nuclear power plants; (2) licensing procedures; (3) safety analysis; (4) project professional services; (5) quality assurance and project organization; (6) construction, scheduling, and operation; (7) nuclear fuel handling and fuel management; (8) plant cost management; and (9) conclusion

The human factors specialist in nuclear control centre design

International Nuclear Information System (INIS)

Wilson, R.B.; Beattie, J.D.

The main focus at Ontario Hydro for man-machine interface design is in the design of control centres. Because the control of a nuclear generating unit is highly centralized there is an increasing need for effective information display and control layout. Control panel design innovations such as the use of CRT displays and the extended use of computerized control in the Darlington station have made it possible for Ontario Hydro to continue to have one first operator for each generating unit. The human factors specialist involved in control panel design must deal with people who know much more about the specific systems being controlled, and must become a generalist in all these systems as well. Designers have to use conceptual techniques such as task analysis, systems design, panel mock-ups, anthropometric data, and personal judgement based on experience as they design panels. They must find a balance between becoming locked into existing technology and methods, slavishly following the latest technological trends, and forgetting that real people will be using what they design

Design quality assurance for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-07-01

This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig.

Design quality assurance for nuclear power plants

International Nuclear Information System (INIS)

1986-07-01

This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig

Analysis and design recommendation on rabbeted capping plate of equipment cell in nuclear chemical facility

International Nuclear Information System (INIS)

Zhang Jingyu; Yin Xiaozhan

2013-01-01

Rabbeted capping plates are widely used in the roof of equipment cells in order to meet the requirements of nuclear radiation protection. The key considerations in the design include vertical load, seismic load and repair load. This article establishes T shaped and Z-shaped plate model via FEM software (ANSYS), analyzes the bearing capacity and displacement distribution in different load cases, and provides recommendations to the design and construction accordingly. (authors)

Stress analysis program system for nuclear vessel: STANSAS

International Nuclear Information System (INIS)

Okamoto, Asao; Michikami, Shinsuke

1979-01-01

IHI has developed a computer system of stress analysis and evaluation for nuclear vessels: STANSAS (STress ANalysis System for Axi-symmetric Structure). The system consists of more than twenty independent programs divided into the following six parts. 1. Programs for opening design by code rule. 2. Calculation model generating programs. 3. Load defining programs. 4. Structural analysis programs. 5. Load data/calculation results plotting programs. 6. Stress evaluation programs. Each program is connected with its pre- or post-processor through three data-bases which enable automatic data transfer. The user can make his choice of structural analysis programs in accordance with the problem to be solved. The interface to STANSAS can be easily installed in generalized structural analysis programs such as NASTRAN and MARC. For almost all tables and figures in the stress report, STANSAS has the function to print or plot out. The complicated procedures of ''Design by Analysis'' for pressure vessels have been well standardized by STANSAS. The system will give a high degree of efficiency and confidence to the design work. (author)

A constraint-based approach to intelligent support of nuclear reactor design

International Nuclear Information System (INIS)

Furuta, Kazuo

1993-01-01

Constraint is a powerful representation to formulate and solve problems in design; a constraint-based approach to intelligent support of nuclear reactor design is proposed. We first discuss the features of the approach, and then present the architecture of a nuclear reactor design support system under development. In this design support system, the knowledge base contains constraints useful to structure the design space as object class definitions, and several types of constraint resolvers are provided as design support subsystems. The adopted method of constraint resolution are explained in detail. The usefulness of the approach is demonstrated using two design problems: Design window search and multiobjective optimization in nuclear reactor design. (orig./HP)

Design and implementation of an MCA card with double ADC for nuclear radiation analysis

International Nuclear Information System (INIS)

Razanabololona, V.H.

2001-01-01

The rapid development of technology has triggered several interesting trends in nuclear instrumentation. In particular, the introduction of add-on cards that can serve as multichannel analyzer and offer adequate performance for most of the nuclear applications. The purpose of this thesis has been the design of a multichannel analyser inserted into microcomputer. The originality of this prototype MCA card with its innovative concept is the use of two analog digital converters or ADC of which the objective is to reduce to minima the lost of pulses during the acquisition and the signals processing. The first ADC works as the main converter (or master) and the second ADC the slave converter. The first ADC begins the first signals acquisition, while it will be busy for the processing, the second ADC continues the acquisition, thus there will be no lost pulses or precisely lost of pulses will be reduced to minima.The three main points for a multichannel analyzer will be studied such a the pile-up, the differential non-linearity or 'DNL' and the dead time. In the first part of this thesis the function of the MCA has been described and the different roles of the composed modules have been developed. In the second part, spectra analysis has been done with the MCA prototype as well the comparison with other MCAs such as MCAs S100, TISA, MCA35+, and the Inspector. [fr

Evaluation of mini super computers for nuclear design applications

International Nuclear Information System (INIS)

Altomare, S.; Baradari, F.

1987-01-01

The evolution of the mini super computers will force changes from the current environment of performing nuclear design calculations on mainframe computers (such as a CRAY) to mini super computers. This change will come about for a number of reasons. First, the mini super computers currently available in the marketplace offer the power and speed comparable to mainframes and can provide the capability to support highly computer intensive calculations. Second, the equipment is physically smaller and can easily be installed and operated without extensive investments in facilities and operations support. Third, the computer capacity can be acquired with as much needed memory, disk, and tape capacity as may be needed. Another reasons is that the performance/cost ratio has increased drastically as hardware costs have decreased. A study was conducted at the Westinghouse Commercial Nuclear Fuel Division (CNFD) to evaluate the mini super computers for use in nuclear core design. As a result of this evaluation, Westinghouse CNFD is offering a combined hardware/software technology transfer package for core design. This package provides the utility designer with a totally dedicated mini super computer comparable in speed to the CRAY 1S with sufficient capacity for a sizable design group to perform the engineering activities related to nuclear core design and operations support. This also assures the utility of being totally compatible with the CNFD design codes, thus assuring total update compatibility

A Critique on the Effectiveness of Current Human Reliability Analysis Approach for the Human-Machine Interface Design in Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Yong Hee

2010-01-01

Human Reliability Analysis (HRA) in cooperation of PSA has been conducted to evaluate the safety of a system and the validity of a system design. HRA has been believed to provide a quantitative value of human error potential and the safety level of a design alternative in Nuclear Power Plants (NPPs). However, it becomes doubtful that current HRA is worth to conduct to evaluate the human factors of NPP design, since there have been many critiques upon the virtue of HRA. Inevitably, the newer the technology becomes, the larger endeavors bound for the new facilitated methods. This paper describes the limitations and the obsolescence of the current HRA, especially for the design evaluation of Human-Machine Interface (HMI) utilizing the recent digital technologies. An alternative approach to the assessment of the human error potential of HMI design is proposed

A cyber security risk assessment for the design of I and C system in nuclear power plants

International Nuclear Information System (INIS)

Song, Jae Gu; Lee, Jung Woon; Lee, Cheal Kwon; Kwon, Kee Choon; Lee, Dong Young

2012-01-01

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the life cycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

A cyber security risk assessment for the design of I and C system in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Song, Jae Gu; Lee, Jung Woon; Lee, Cheal Kwon; Kwon, Kee Choon; Lee, Dong Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-12-15

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the life cycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

Requests on domestic nuclear data library from BWR design

International Nuclear Information System (INIS)

Maruyama, Hiromi

2003-01-01

Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

Improved design on Qinshan 300 MWe nuclear power plant

International Nuclear Information System (INIS)

Shi Peihua; Cheng Wanli; Lu Rongliang

1993-01-01

The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

Improved design on Qinshan 300 MWe nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

1993-06-01

The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

LMFBR core design analysis

International Nuclear Information System (INIS)

Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

1980-01-01

The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

Revolution of Nuclear Power Plant Design Through Digital Technology

International Nuclear Information System (INIS)

Zhang, L.; Shi, J.; Chen, W.

2015-01-01

In the digital times, digital technology has penetrated into every industry. As the highest safety requirement standard, nuclear power industry needs digital technology more to breed high quality and efficiency. Digital power plant is derived from digital design and the digitisation of power plant transfer is an inevitable trend. This paper introduces the technical solutions and features of digital nuclear power plant construction by Shanghai Nuclear Engineering Research & Design Institute, points out the key points and technical difficulties that exist in the process of construction and can serve as references for further promoting construction of digital nuclear power plant. Digital technology is still flourishing. Although many problems will be encountered in construction, it is believed that digital technology will make nuclear power industry more safe, cost-effective and efficient. (author)

Seismic analysis of the Aguirre Nuclear Reactor

International Nuclear Information System (INIS)

Sepulveda Soza, Cristian

1999-01-01

This thesis aims to verify the seismic design of the Aguirre Nuclear Reactor using the finite elements method and comparing the results with the original analysis. The study focused on the dynamic interaction of soil and structures, using the ANSYS program for the analysis, which was implemented for a work station under a UNIX platform belonging to the Chilean Nuclear Energy Commission. The modeling of the structures was carried out following International Atomic Energy recommendations, those of the makers of the Swanson Analysis Systems program and the prior study by S y S Ingenieros Consultores. Two-dimensional models were developed with axial and symmetry and three-dimensional models with symmetric and asymmetric plans, where the retaining building, the pond block and the soil down to the basal rock were included. The seismic stresses were defined according to the Chilean Standard NCh433.of96, using the spectrum of design accelerations for type II soils for the structural models and type IV for the soil-structure interaction models.The results of interest for this study are: the compression and cutting tensions, the unitary cut distortions and the displacements, which are shown graphically and are compared between the different models and with the original analysis. A sensitivity analysis was prepared for the models with axial symmetry considering soil reaction coefficient values of 20, 10, 5, 2, 1 and 0.5 kp/cm 3 ; and four screens with maximum sizes of 100, 50, 25 and 12.5 cm. The behavior of the stressed materials was studied as well as the result of the seismic stress (CS)

Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Framework for AI-based nuclear reactor design support system

International Nuclear Information System (INIS)

Furuta, Kazuo; Kondo, Shunsuke

1992-01-01

Nowadays many computer programs are being developed and used for the analytic tasks in nuclear reactor design, but experienced designers are still responsible for most of the synthetic tasks which are not amenable to algorithmic computer processes. Artificial intelligence (AI) is a promising technology to deal with these intractable tasks in design. In development of AI-based design support systems, it is desirable to choose a comprehensive framework based on the scientific theory of design. In this work a framework for AI-based design support systems for nuclear reactor design will be proposed based on an explorative abduction model of design. The fundamental architectures of this framework will be described especially on knowledge representation, context management and design planning. (author)

Seismic design of equipment and piping systems for nuclear power plants in Japan

International Nuclear Information System (INIS)

Minematsu, Akiyoshi

1997-01-01

The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

Seismic design of equipment and piping systems for nuclear power plants in Japan

Energy Technology Data Exchange (ETDEWEB)

Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

1997-03-01

The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

The design of nuclear radiation measuring instrument of embedded network

International Nuclear Information System (INIS)

Zhang Huaiqiang; Ge Liangquan; Xiong Shengqing

2009-01-01

The design and realization of nuclear radiation measuring instrument is introduced. Due to the current nuclear instrument often used serial interface to communicate the PC, it is widely used for simple design and easy operation. However, as the demand of remote data acquisition and the call of sharing resources, the design of embedded the TCP/IP protocol stack into MCU, it may send the nuclear signal in Internet. Some devices that link each other with the network can be networked. The design is not only realizing remote data acquisition and sharing resources, but also reducing costs and improving the maintainability of the system. (authors)

Seismic response analysis for a deeply embedded nuclear power plant

International Nuclear Information System (INIS)

Chen, W.W.H.; Chatterjee, M.; Day, S.M.

1979-01-01

One of the important aspect of the aseimic design of nuclear power plants is the evaluation of the seismic soil-structure interaction effect due to design earthquakes. The soil-structure interaction effect can initiate rocking and result in different soil motions compared to the free field motions, thus significantly affecting the structural response. Two methods are generally used to solve the seismic soil-structure interaction problems: the direct finite element method (FLUSH) and the substructure or impedance approach. This paper presents the results of the horizontal seismic soil-structure interaction analysis using the impedance aproach and the direct finite element method for a deeply embedded nuclear power plant. (orig.)

Conceptual design of light ion beam inertia nuclear fusion reactors

International Nuclear Information System (INIS)

1983-07-01

Light ion beam, inertia nuclear fusion system drew attention recently as one of the nuclear fusion systems for power reactors in the history of the research on nuclear fusion. Its beginning seemed to be the judgement that the implosion of fusion fuel pellets with light ions can be realized with the light ions which can be obtained in view of accelerator techniques. Of course, in order to generate practically usable nuclear fusion reaction by this system and maintain it, many technical difficulties must be overcome. This research was carried out for the purpose of discovering such technical problems and searching for their solution. At the time of doing the works, the following policy was adopted. Though their is the difference of fine and rough, the design of a whole reactor system is performed conformably. In order to make comparison with other reactor types and nuclear fusion systems, the design is carried out as the power plant of about one million kWe output. As the extent of the design, the works at conceptual design stage are performed to present the concept of design which satisfies the required function. Basically, the design is made from conservative standpoint. This research of design was started in 1981, and in fiscal 1982, the mutual adjustment among the design of respective parts was performed on the basis of the results in 1981, and the possible revision and new proposal were investigated. (Kako, I.)

Summary of Prometheus Radiation Shielding Nuclear Design Analyses , for information

International Nuclear Information System (INIS)

J. Stephens

2006-01-01

This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL and Bettis) shielding nuclear design analyses done for the project

Economic analysis of KNGR's conceptual design

International Nuclear Information System (INIS)

Roh, Myung-Sub; Chung, Kyung-Nam; Hong, Jang-Hee

1997-01-01

During the last two decades in Korea, with the rapid economic growth and industrialization, nuclear power has played an important role in electric power production. At present, Korea has 11 nuclear units in operation with total installed capacity of 9,916MWe and 7 units under construction with total capacity of 6,100MWe. An advanced reactor development program, called Korean Next Generation Reactor (KNGR), has been started for the coming 21st century. It aims to enhance safety features compared to existing plants and follow the target to maintain the economic competitiveness of nuclear energy with alternative energy sources. In order to meet these requirements, a number of technical and economical factors have been taken into the program. These factors include the economy of plant size, design simplification and optimization, and reduction of construction period. This paper addresses the preliminary economic analysis results for the KNGR design. It is indicated that KNGR has about 17% economic advantage compared to the current 1,000 MWe PWR. This paper also describes some of the experiences gained and important factors related to reducing the investment and operation costs. 5 refs., 2 figs., 5 tabs

Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

2014-06-15

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (k{sub eff}), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

Role of civil engineers in design and construction of nuclear power plants

International Nuclear Information System (INIS)

Sivathanu Pillai, C.

2015-01-01

Nuclear energy remains an important element in India's energy mix for sustaining rapid economic growth. In the Indian energy scenario projections for the future, nuclear power through fast reactors has been considered as an important component. Fast reactors are also planned to be built in the medium term to burn minor actinides as means of reducing the burden of radioactive waste management. Nuclear power in India provides 3% of the electricity generated in the country now; it is estimated to go up to 25% by 2050. India has a special interest in developing fast breeder reactors and use of thorium as a source of energy as it has one of the largest reserves of thorium. Nuclear industry always provides challenging works to civil engineers. From geotechnical investigation, structural analysis, design, construction and to material developments and provides unique opportunity for excellence

JAERI thermal reactor standard code system for reactor design and analysis SRAC

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1985-01-01

SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

FAULT PROPAGATION AND EFFECTS ANALYSIS FOR DESIGNING AN ONLINE MONITORING SYSTEM FOR THE SECONDARY LOOP OF A NUCLEAR POWER PLANT PART OF A HYBRID ENERGY SYSTEM

Energy Technology Data Exchange (ETDEWEB)

Li, Huijuan; Diao, Xiaoxu; Li, Boyuan; Smidts, Carol; Bragg-Sitton, Shannon

2017-03-01

This paper studies the propagation and effects of faults of critical components that pertain to the secondary loop of a nuclear power plant found in Nuclear Hybrid Energy Systems (NHES). This information is used to design an on-line monitoring (OLM) system which is capable of detecting and forecasting faults that are likely to occur during NHES operation. In this research, the causes, features, and effects of possible faults are investigated by simulating the propagation of faults in the secondary loop. The simulation is accomplished by using the Integrated System Failure Analysis (ISFA). ISFA is used for analyzing hardware and software faults during the conceptual design phase. In this paper, the models of system components required by ISFA are initially constructed. Then, the fault propagation analysis is implemented, which is conducted under the bounds set by acceptance criteria derived from the design of an OLM system. The result of the fault simulation is utilized to build a database for fault detection and diagnosis, provide preventive measures, and propose an optimization plan for the OLM system.

Radiochemistry and nuclear methods of analysis

International Nuclear Information System (INIS)

Ehmann, W.D.; Vance, D.

1991-01-01

This book provides both the fundamentals of radiochemistry as well as specific applications of nuclear techniques to analytical chemistry. It includes such areas of application as radioimmunoassay and activation techniques using very short-lined indicator radionuclides. It emphasizes the current nuclear methods of analysis such as neutron activation PIXE, nuclear reaction analysis, Rutherford backscattering, isotope dilution analysis and others

Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

PIEPHO, M.G.

1999-10-20

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

Comprehensive evaluation method in application of nuclear DCS product design

International Nuclear Information System (INIS)

Wang Weixin; Zhao Zhemin; Shi Yingbin

2014-01-01

In order to select the best design proposal in short time, the TOPSIS comprehensive evaluation method in the nuclear power plant DCS product design was introduced. It can intuitively show the different design proposals good or not good by data and shorten the time of the design proposal optimization. The design proposal selected by this method will be more reasonable and has good comprehensive performance indexes. The TOPSIS comprehensive evaluation method achieves good result in one of the nuclear power plant DCS cabinet design proposal optimization. (authors)

Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

2007-03-01

One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database

Design and verification of computer-based reactor control system modification at Bruce-A candu nuclear generating station

International Nuclear Information System (INIS)

Basu, S.; Webb, N.

1995-01-01

The Reactor Control System at Bruce-A Nuclear Generating Station is going through some design modifications, which involve a rigorous design process including independent verification and validation. The design modification includes changes to the control logic, alarms and annunciation, hardware and software. The design (and verification) process includes design plan, design requirements, hardware and software specifications, hardware and software design, testing, technical review, safety evaluation, reliability analysis, failure mode and effect analysis, environmental qualification, seismic qualification, software quality assurance, system validation, documentation update, configuration management, and final acceptance. (7 figs.)

FPGA-based technology for Pulse Height Analysis in nuclear spectrometry system

International Nuclear Information System (INIS)

Andrianiaina, H.; Raoelina Andriambololona; Rajaobelison, J.; Rambolamanana, G.; Roengen, H.

2009-01-01

Facing the rapid technology development applied in nuclear instruments, Madagascar-INSTN has taken measures to estabilish and enhance the national capabilities in their maintenance, calibration, design/modification and repair, which is a key factor in the development of sustainable nuclear technology.The maintenance will be more convenient if all technical resources are made available: this is only possible when the acces to designer side of the instruments is permitted. Research and instrumentation development projects have been developed to support and to fit the local needs: Pulse Height Analysis algorithm is designed within FPGA and VHDL programming technics to build a Multi-Input Multi-Channel Analyzer. Description and measurement results with the MIMCA will be described in the present paper

Ultimate capacity and influenced factors analysis of nuclear RC containment subjected to internal pressure

International Nuclear Information System (INIS)

Song Chenning; Hou Gangling; Zhou Guoliang

2014-01-01

Ultimate compressive bearing capacity, influenced factors and its rules of nuclear RC containment are key problems of safety assessment, accident treatment and structure design, etc. Ultimate compressive bearing capacity of nuclear RC containment is shown by concrete damaged plasticity model and steel double liner model of ABAQUS. The study shows that the concrete of nuclear RC containment cylinder wall becomes plastic when the internal pressure is up to 0.87 MPa, the maximum tensile strain of steel liner exceeds 3000 × 10 6 and nuclear RC containment reaches ultimate status when the internal pressure is up to 1.02 MPa. The result shows that nuclear RC containment is in elastic condition under the design internal pressure and the bearing capacity meets requirement. Prestress and steel liner play key parts in the ultimate internal pressure and failure mode of nuclear RC containment. The study results have value for the analysis of ultimate compressive bearing capacity, structure design and safety assessment. (authors)

Human factors design guidelines for maintainability of Department of Energy nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Bongarra, J.P. Jr.; VanCott, H.P.; Pain, R.F.; Peterson, L.R.; Wallace, R.I.

1985-06-18

Intent of these guidelines is to provide design and design review teams of DOE nuclear facilities with human factors principles to enhance the design and aid in the inspection of DOE nuclear facilities, systems, and equipment. These guidelines are concerned with design features of DOE nuclear facilities which can potentially affect preventive and corrective maintenance of systems within DOE nuclear facilities. Maintenance includes inspecting, checking, troubleshooting, adjusting, replacing, repairing, and servicing activities. Other factors which influence maintainability such as repair and maintenance suport facilities, maintenance information, and various aspects of the environment are also addressed.

Human factors design guidelines for maintainability of Department of Energy nuclear facilities

International Nuclear Information System (INIS)

Bongarra, J.P. Jr.; VanCott, H.P.; Pain, R.F.; Peterson, L.R.; Wallace, R.I.

1985-01-01

Intent of these guidelines is to provide design and design review teams of DOE nuclear facilities with human factors principles to enhance the design and aid in the inspection of DOE nuclear facilities, systems, and equipment. These guidelines are concerned with design features of DOE nuclear facilities which can potentially affect preventive and corrective maintenance of systems within DOE nuclear facilities. Maintenance includes inspecting, checking, troubleshooting, adjusting, replacing, repairing, and servicing activities. Other factors which influence maintainability such as repair and maintenance suport facilities, maintenance information, and various aspects of the environment are also addressed

Nuclear analysis

International Nuclear Information System (INIS)

1988-01-01

Basic studies in nuclear analytical techniques include the examination of underlying assumptions and the development and extention of techniques involving the use of ion beams for elemental and mass analysis. 1 ref., 1 tab

Design and axial optimization of nuclear fuel for BWR reactors

International Nuclear Information System (INIS)

Garcia V, M.A.

2006-01-01

In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

Building Foundations for Nuclear Security Enterprise Analysis Utilizing Nuclear Weapon Data

Energy Technology Data Exchange (ETDEWEB)

Josserand, Terry Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis; Young, Leone [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis; Chamberlin, Edwin Phillip [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis

2017-09-01

The Nuclear Security Enterprise, managed by the National Nuclear Security Administration - a semiautonomous agency within the Department of Energy - has been associated with numerous assessments with respect to the estimating, management capabilities, and practices pertaining to nuclear weapon modernization efforts. This report identifies challenges in estimating and analyzing the Nuclear Security Enterprise through an analysis of analogous timeframe conditions utilizing two types of nuclear weapon data - (1) a measure of effort and (2) a function of time. The analysis of analogous timeframe conditions that utilizes only two types of nuclear weapon data yields four summary observations that estimators and analysts of the Nuclear Security Enterprise will find useful.

Economic analysis of nuclear energy

Energy Technology Data Exchange (ETDEWEB)

Lee, Han Myung; Lee, M.K.; Moon, K.H.; Kim, S.S.; Lim, C.Y.; Song, K.D.; Kim, H

2001-12-01

The objective of this study is to evaluate the contribution of nuclear energy to the energy use in the economical way, based on the factor survey performed on the internal and external environmental changes occurred recent years. Internal and external environmental changes are being occurred recent years involving with using nuclear energy. This study summarizes the recent environmental changes in nuclear energy such as sustainable development issues, climate change talks, Doha round and newly created electricity fund. This study also carried out the case studies on nuclear energy, based on the environmental analysis performed above. The case studies cover following topics: role of nuclear power in energy/environment/economy, estimation of environmental external cost in electric generation sector, economic comparison of hydrogen production, and inter-industrial analysis of nuclear power generation.

Economic analysis of nuclear energy

International Nuclear Information System (INIS)

Lee, Han Myung; Lee, M.K.; Moon, K.H.; Kim, S.S.; Lim, C.Y.; Song, K.D.; Kim, H.

2001-12-01

The objective of this study is to evaluate the contribution of nuclear energy to the energy use in the economical way, based on the factor survey performed on the internal and external environmental changes occurred recent years. Internal and external environmental changes are being occurred recent years involving with using nuclear energy. This study summarizes the recent environmental changes in nuclear energy such as sustainable development issues, climate change talks, Doha round and newly created electricity fund. This study also carried out the case studies on nuclear energy, based on the environmental analysis performed above. The case studies cover following topics: role of nuclear power in energy/environment/economy, estimation of environmental external cost in electric generation sector, economic comparison of hydrogen production, and inter-industrial analysis of nuclear power generation

Development of COMPAS, computer aided process flowsheet design and analysis system of nuclear fuel reprocessing

International Nuclear Information System (INIS)

Homma, Shunji; Sakamoto, Susumu; Takanashi, Mitsuhiro; Nammo, Akihiko; Satoh, Yoshihiro; Soejima, Takayuki; Koga, Jiro; Matsumoto, Shiro

1995-01-01

A computer aided process flowsheet design and analysis system, COMPAS has been developed in order to carry out the flowsheet calculation on the process flow diagram of nuclear fuel reprocessing. All of equipments, such as dissolver, mixer-settler, and so on, in the process flowsheet diagram are graphically visualized as icon on a bitmap display of UNIX workstation. Drawing of a flowsheet can be carried out easily by the mouse operation. Not only a published numerical simulation code but also a user's original one can be used on the COMPAS. Specifications of the equipment and the concentration of components in the stream displayed as tables can be edited by a computer user. Results of calculation can be also displayed graphically. Two examples show that the COMPAS is applicable to decide operating conditions of Purex process and to analyze extraction behavior in a mixer-settler extractor. (author)

Design and qualification of nuclear fittings

International Nuclear Information System (INIS)

Coppolani, P.

1985-01-01

This paper gives a definition of nuclear fittings and presents its characteristics, its operating conditions and the safety function that they ensure. The classification of fittings and the consequences on its design are presented. The procedures of design verification and the validation in loop are given. Two equipments are finally presented: the Framatome Velan Rateau valve and the VR3-1500 LB valve [fr

An AI-based layout design system for nuclear power plants

International Nuclear Information System (INIS)

Fujita, Kikuo; Akagi, Shinsuke; Nakatogawa, Tetsundo; Tanaka, Kazuo; Takeuchi, Makoto.

1991-01-01

An AI-based layout design system for nuclear power plants has been developed. The design of the layout of nuclear power plants is a time-consuming task requiring expertise, in which a lot of machinery and equipment must be arranged in a plant building considering various kinds of design constraints, i.e. spatial, functional, economical etc. Computer aided layout design systems have been widely expected and the application of AI technology is expected as a promising approach for the synthesis phase of this task. In this paper, we present an approach to the layout design of nuclear power plants based on a constraint-directed search; one of the AI techniques. In addition, we show how it was implemented with an object-oriented programming technique and give an example of its application. (author)

Detailed 3-D nuclear analysis of ITER outboard blanket modules

Energy Technology Data Exchange (ETDEWEB)

Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

2015-10-15

Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

Detailed 3-D nuclear analysis of ITER outboard blanket modules

International Nuclear Information System (INIS)

Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

2015-01-01

Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

Information management for nuclear power stations: System Design Concept

International Nuclear Information System (INIS)

Halpin, D.W.

1978-03-01

A study of the information management structure required to support nuclear power plant construction was performed by a joint university-industry group under the sponsorship of the Department of Energy (DOE), formerly the Energy Research and Development Administration (ERDA). The purpose of this study was (1) to study methods for the control of information during the construction and start-up of nuclear power plants, and (2) identify those data elements intrinsic to nuclear power plants which must be maintained in a structured format for quick access and retrieval. Maintenance of the massive amount of data needed for control of a nuclear project during design, procurement, construction, start-up/testing, and operational phases requires a structuring which allows immediate update and retrieval based on a wide variety of access criteria. The objective of the research described has been to identify design concepts which support the development of an information control system responsive to these requirements. A conceptual design of a Management Information Data Base System which can meet the project control and information exchange needs of today's large nuclear power plant construction projects has been completed and an approach recommended for development and implementation of a complete operational system

Structural analysis of aircraft impact on a nuclear powered ship

International Nuclear Information System (INIS)

Dietrich, R.

1976-01-01

As part of a general safety analysis, the reliability against structural damage due to an aircraft crash on a nuclear powered ship is evaluated. This structural analysis is an aid in safety design. It is assumed that a Phantom military jet-fighter hits a nuclear powered ship. The total reaction force due to such an aircraft impact on a rigid barrier is specified in the guidelines of the Reaktor-Sicherheitskommission (German Safety Advisory Committee) for pressurized water reactors. This paper investigates the aircraft impact on the collision barrier at the side of the ship. The aircraft impact on top of the reactor hatchway is investigated by another analysis. It appears that the most unfavorable angle of impact is always normal to the surface of the collision barrier. Consequently, only normal impact will be considered here. For the specific case of an aircraft striking a nuclear powered ship, the following two effects are considered: Local penetration and dynamic response of the structure. (Auth.)

New design system for nuclear power plant

International Nuclear Information System (INIS)

Kakuta, Masataka; Yoshinaga, Toshiaki; Yoshida, Ikuzo; Tokumasu, Shinji.

1980-01-01

As for the machine and equipment layout and the piping design for nuclear power plants, the multilateral coordination and study on such factors as functions, installation, radiation exposure and maintenance are required, and the high reliability is demanded. On the other hand, the quantity of things handled is enormous, therefore it is difficult to satisfy completely the above described requirements and to make plant planning which is completely free from the mutual interference of machines, equipments and pipings by the ordinary design with drawings only. Thereupon, the following new device was adopted to the design method for the purposes of improving the quality and shortening the construction period. Namely at the time of designing new plants, the rationalization of plant planning method was attempted by introducing color composite drawings and the technique of model engineering, at the same time, the newly developed design system for pipings was applied with a computer, thus the large accomplishment was able to be obtained regarding the improvement of reliability and others by making the check-up of the propriety. The design procedures of layout and piping, the layout design and general coordination in nuclear power stations with models and color composite drawings and the design system are explained. (Kako, I.)

Safety criteria for design of nuclear power plants

International Nuclear Information System (INIS)

1997-01-01

In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

Job analysis for training design and evaluation

International Nuclear Information System (INIS)

Andersson, H.; Baeck, P.; Wirstad, J.

1979-01-01

The report describes a job analysis method which has been used for competency requirements and training evaluation purposes on three jobs in a Swedish nuclear power station. The jobs are the shift supervisor, the reactor operator and the turbine operator. The method is based on a system analytical approach. Through a topdown analysis of the power generation system a structure is derived which is used for an inventory of possible operator tasks. The primary source for actual task generation was interviews with operators and with other personnel who are cooperating with the operators. The method suggests alternative sources for the task generation when operator interviews are not feasible, e.g. in systems which are being developed or when operator experiences are lacking. The result of the job analysis is a set of typical tasks for each job. These tasks represent the contents of the job to be considered in the planning of job training and for recruitment requirements. The same job analysis method was used for training design and evaluation in a Finnish nuclear power station. The concluding remarks of this report are made in the light of this study also. This report is based on a research project B15/77 from the Swedish Nuclear Power Inspectorate. (author)

Summary report of workshop on sabotage protection in nuclear power plant design

International Nuclear Information System (INIS)

1977-02-01

During the Summer of 1976, Sandia Laboratories hosted a workshop on Sabotage Protection in Nuclear Power Plant Design in which 11 consultants from the nuclear power industry participated. Each consultant is highly qualified and experienced in nuclear power plant design. The objective of the workshop was to identify practicable design measures which could be employed in future nuclear power plants to provide increased protection against sabotage. The report summarizes the conclusions and recommendations of the workshop

Design Lessons Drawn from the Decommissioning of Nuclear Facilities

International Nuclear Information System (INIS)

2011-05-01

This report provides an updated compilation incorporating the most recent lessons learned from decommissioning and remediation projects. It is intended as a 'road map' to those seeking to apply these lessons. The report presents the issues in a concise and systematic manner, along with practical, thought-provoking examples. The most important lessons learned in recent years are organized and examined to enable the intended audience to gauge the importance of this aspect of the planning for new nuclear facilities. These will be of special interest to those seeking to construct nuclear facilities for the first time. In Sections 1 and 2, the current situation in the field of decommissioning is reviewed and the relevance and importance of beneficial design features is introduced. A more detailed review of previous and current lessons learned from decommissioning is given in Section 3 where different aspects of the decommissioning process are analysed. From this analysis beneficial design features have been extracted and identified in Section 4 which includes two comprehensive tables where brief descriptions of the features are summarized and responsibilities are identified. Conclusions and key design features and key recommendations are given in Section 5. Two Annexes are included to provide lessons from past projects and past experience and to record notes and extracts taken from a comprehensive list of publications listed in the References on page 47.

LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

International Nuclear Information System (INIS)

Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.