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Sample records for nuclear design analysis

  1. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  2. 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

  3. Linear regression and sensitivity analysis in nuclear reactor design

    International Nuclear Information System (INIS)

    Kumar, Akansha; Tsvetkov, Pavel V.; McClarren, Ryan G.

    2015-01-01

    Highlights: • Presented a benchmark for the applicability of linear regression to complex systems. • Applied linear regression to a nuclear reactor power system. • Performed neutronics, thermal–hydraulics, and energy conversion using Brayton’s cycle for the design of a GCFBR. • Performed detailed sensitivity analysis to a set of parameters in a nuclear reactor power system. • Modeled and developed reactor design using MCNP, regression using R, and thermal–hydraulics in Java. - Abstract: The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data

  4. Analysis and design of nuclear energy information systems

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Sriyana; Arief Tris Yuliyanto; Wiku Lulus Widodo

    2015-01-01

    Management of research reports and activities of the Center for Nuclear Energy System Assessment (PKSEN), either in the form of documents and the results of other activities, are important part of the series of activities PKSEN mission achievement. Management of good documents will facilitate the provision of improved inputs or use the maximum results. But over the past few years, there are still some problem in the management of research reports and activities performed by PKSEN. The purpose of this study is to analyze and design flow layout of the Nuclear Energy Information System to facilitate the implementation of the Nuclear Energy Information System. In addition to be used as a research management system and PKSEN activities, it can also be used as information media for the community. Nuclear Energy Information System package is expected to be ''one gate systems for PKSEN information. The research methodology used are: (i) analysis of organizational systems, (ii) the analysis and design of information systems; (iii) the analysis and design of software systems; (iv) the analysis and design of database systems. The results of this study are: had identified and resources throughout the organization PKSEN activation, had analyzed the application of SIEN using SWOT analysis, had identified several types of devices required, had been compiled hierarchy of SIEN, had determined that the database system used is a centralized database system and had elections MySQL as DBMS. The result is a basic design of the Nuclear Energy Information System) which will used as a research and activities management system of PKSEN and also can be used as a medium of information for the community. (author)

  5. Development of design and analysis software for advanced nuclear system

    International Nuclear Information System (INIS)

    Wu Yican; Hu Liqin; Long Pengcheng; Luo Yuetong; Li Yazhou; Zeng Qin; Lu Lei; Zhang Junjun; Zou Jun; Xu Dezheng; Bai Yunqing; Zhou Tao; Chen Hongli; Peng Lei; Song Yong; Huang Qunying

    2010-01-01

    A series of professional codes, which are necessary software tools and data libraries for advanced nuclear system design and analysis, were developed by the FDS Team, including the codes of automatic modeling, physics and engineering calculation, virtual simulation and visualization, system engineering and safety analysis and the related database management etc. The development of these software series was proposed as an exercise of development of nuclear informatics. This paper introduced the main functions and key techniques of the software series, as well as some tests and practical applications. (authors)

  6. Comparative analysis of nuclear reactor control system designs

    International Nuclear Information System (INIS)

    Russcher, G.E.

    1975-01-01

    Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

  7. Optimizing Nuclear Reaction Analysis (NRA) using Bayesian Experimental Design

    International Nuclear Information System (INIS)

    Toussaint, Udo von; Schwarz-Selinger, Thomas; Gori, Silvio

    2008-01-01

    Nuclear Reaction Analysis with 3 He holds the promise to measure Deuterium depth profiles up to large depths. However, the extraction of the depth profile from the measured data is an ill-posed inversion problem. Here we demonstrate how Bayesian Experimental Design can be used to optimize the number of measurements as well as the measurement energies to maximize the information gain. Comparison of the inversion properties of the optimized design with standard settings reveals huge possible gains. Application of the posterior sampling method allows to optimize the experimental settings interactively during the measurement process.

  8. Design analysis and microprocessor based control of a nuclear reactor

    International Nuclear Information System (INIS)

    Sabbakh, N.J.

    1988-01-01

    The object of this thesis is to design and test a microprocessor based controller, to a simulated nuclear reactor system. The mathematical model that describes the dynamics of a typical nuclear reactor of one group of delayed neutrons approximations with temperature feedback was chosen. A digital computer program has been developed for the design and analysis of a simulated model based on the concept of state-variable feedback in order to meet a desired system response with maximum overshoot of 3.4% and setting time of 4 sec. The state variable feedback coefficients are designed for the continuous system, then an approximation is used to obtain in the state variable feedback vector for the discrete system. System control was implemented utilizing Direct Digital Control (DDC) of a nuclear reactor simulated model through a control algorithm that was performed by means of a microprocessor based system. The controller performance was satisfactorily tested by exciting the reactor system with a transient reactivity disturbance and by a step change in power demand. Direct digital control, when implemented on a microprocessor adds versatility, flexibility in system design with the added advantage of possible use of optimal control algorithms. 6 tabs.; 30 figs.; 46 refs.; 6 apps

  9. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    International Nuclear Information System (INIS)

    Chen, Shucheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at the NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance parameters of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed

  10. 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.)

  11. 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

  12. 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)

  13. 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

  14. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  15. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-15

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  16. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  17. Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

    International Nuclear Information System (INIS)

    Widargo, Reza; Anghaie, Samim

    1999-01-01

    The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight

  18. Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

    International Nuclear Information System (INIS)

    Goto, Minoru

    2015-03-01

    An appropriate configuration of fuel and reactivity control equipment in a nuclear reactor core, which allows the design of the nuclear reactor core for low cost and high performance, is performed by nuclear design with high accuracy. The accuracy of nuclear design depends on a nuclear data library and a nuclear analysis method. Additionally, it is one of the most important issues for the nuclear design of a High Temperature Gas-cooled Reactor (HTGR) that an insertion depth of control rods into the reactor core should be retained shallow by reducing excess reactivity with a different method to keep fuel temperature below its limitation thorough a burn-up period. In this study, using experimental data of the High Temperature engineering Test Reactor (HTTR), which is a Japan's HTGR with 30 MW of thermal power, the following issues were investigated: applicability of nuclear data libraries to nuclear analysis for HTGRs; applicability of the improved nuclear analysis method for HTGRs; and effectiveness of a rod-type burnable poison on HTGR reactivity control. A nuclear design of a small-sized HTGR with 50 MW of thermal power (HTR50S) was performed using these results. In the nuclear design of the HTR50S, we challenged to decrease the kinds of the fuel enrichments and to increase the power density compared with the HTTR. As a result, the nuclear design was completed successfully by reducing the kinds of the fuel enrichment to only three from twelve of the HTTR and increasing the power density by 1.4 times as much as that of the HTTR. (author)

  19. Nuclear power plant analysers: their approach to analysis and design

    International Nuclear Information System (INIS)

    Ancarani, A.; Zanobetti, D.

    1985-01-01

    ''Analysers'' as used for nuclear power plant simulators are powerful tools and their purpose can be variously assigned: it may vary from the aid in the design of power plants to the assistance to operators in emergency situations. A fundamental problem arising from the analysers' concept and use is the definition of the simulation capability. This can be assessed either by comparison with previous operational data statistically significant and suitably elaborated; or by comparison with theoretical (computed) values obtained from engineering codes. In both these, to take advantage of all the possibilities offered by the ''analysers'', it is mandatory that suitable terms of reference be clearly stated and agreed upon. Particular care is devoted to accuracy in the prediction of physical values both for the steady state and the transient situations. For instance, it can be seen that such evaluations can be met by specifying the maximum error on value of parameters (ordinates), save for very fast transients; the maximum error on time (abscissae) for occurrence of extreme values; the maximum error on values of extremes (ordinates); the maximum error on derivatives (slopes) for rapidly variable transients, save near extreme values. The paper also deals with a brief account of the present projects and proposals in different countries as known from various sources, and mentions a possible co-ordination at international level. (author)

  20. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  1. Seismic design and analysis of nuclear fuel cycle facilities in France

    International Nuclear Information System (INIS)

    Sollogoub, P.

    2001-01-01

    Methodology for seismic design of nuclear fuel facilities and power plants in France is described. After the description of regulatory and normative texts for seismic design, different elements are examined: definition of ground motion, analysis methods, new trends, reevaluation and specificity of Fuel Cycle Facilities. R/D developments are explicated in each part. Their final objective are to better quantify the margins of each step which, in relation with safety analysis,lead to balanced design, analysis and retrofit rules. (author)

  2. Design and analysis challenges for advanced nuclear fuel

    International Nuclear Information System (INIS)

    Klepfer, H.; Abdollahian, D.; Dias, A.; Durston, C.; Eisenhart, L.; Engel, R.; Gilmore, P.; Rank, P.; Kjaer-Pedersen, N.; Sorensen, J.; Yang, R.; Agee, L.

    2004-01-01

    Significant changes have been incorporated in the light water reactor (LWR) fuel designs now being offered, and advanced fuel designs are currently being developed for the existing and the next generation of reactor designs. These advanced fuel design configurations are intended to offer utilities major economic gains, including: (1) improved fuel characteristics through optimized hydrogen to uranium ratio within the core; (2) increased capacity factor by allowing longer operating cycles, which is implemented by increasing the fuel enrichment and the amount and distribution of burnable poison, gadolinia, boron, or erbium within the fuel assembly to achieve higher discharge burnup; and (3) increased plant power output, if it can be accommodated by the balance of plant, by increasing the power density of the fuel assembly. The authors report here work being done to identify emerging technical issues in support of utility industry evaluations of advanced fuel designs. (author)

  3. Dynamic analysis and structural design of underground nuclear reactor containments

    International Nuclear Information System (INIS)

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

    1975-01-01

    All concept options are assumed to be similar in design criteria for structural competence to contain radioactivity and fuel heat and meet the functional, servicing, protective and aesthetic requirements. The choice of underground siting should be based on criteria developed from the sequential consideration of load-causing phenomena, concept and site characteristics. From the criteria, loads for a particular concept and site are calculated and the design formulated. (orig./ORU) [de

  4. 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

  5. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  6. 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.)

  7. 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)

  8. Advanced concepts, analysis approaches and criteria for nuclear piping system design

    International Nuclear Information System (INIS)

    Tang, H.T.; Tagart, S.W. Jr.; Tang, Y.K.

    1992-01-01

    Recent research in piping system design and analysis has resulted in advancements on damping values, independent support motion (ISM), static coefficient method, simplified inelastic method and ASME code criteria changes. In the support area, passive type of supports such as energy-absorbing device and gap stopper have been developed. These advancements provide bases for improved and cost-effective design of future nuclear piping systems. (author)

  9. Implementation of the Westinghouse nuclear design system for incore fuel management analysis

    International Nuclear Information System (INIS)

    Hoskins, K.C.; Kichty, M.J.; Liu, Y.S.; Nguyen, T.Q.

    1990-01-01

    Development of the Westinghouse Advanced Nuclear Design System, which includes PHOENIX-P and ANC, has been continued to improve the efficiency, reliability, accuracy, and flexibility of models. The new codes ALPHA and PHIRE provide complete automation and interface functions for PHOENIX-P, ANC, and other codes. PHOENIX-P has been modified to generate data for ANC based on single or multi-assembly calculations. ANC has several enhancements, including improved pin power reconstruction, automated 2D model generation, and rod burnup prediction capability. The excellent performance of PHOENIX-P/ANC models is demonstrated by the results of over 30 models covering the range of Westinghouse designs. This Nuclear Design System is now the standard Westinghouse methodology for core design and analysis

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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)

  15. Ultrasonic field analysis program for transducer design in the nuclear industry

    International Nuclear Information System (INIS)

    Singh, G.P.; Rose, J.L.

    1980-02-01

    An ultrasonic field analysis program is presented that can be used for transducer design in the nuclear industry. Calculation routines that make use of Huygen's principle in a field analysis model are introduced that enable such field characteristics as axial and lateral resolution, beam symmetry, and gain variation throughout the ultrasonic field to be optimized. Mathematical details are presented along with several sample problems that show comparisons with classical results reported in the literature and with experimental data. Several sample problems that are of interest to the nuclear industry are also included, along with some that satisfy both academical and practical curiosity. These include transducer shape effects, pulse shape effects, crystal vibration variation, and an introduction to such novel transducer designs as annular arrays and dual element angle beam transducers

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. Problems of structural mechanics in nuclear design

    International Nuclear Information System (INIS)

    Patwardhan, V.M.; Kakodkar, Anil

    1975-01-01

    A very careful and detailed stress analysis of nuclear presure vessels and components is essential for ensuring the safety and integrity of nuclear power plants. The nuclear designer, therefore, relies heavily on structural mechanics for application of the most advanced stress analysis techniques to practical design problems. The paper reviews the inter-relation between structural mechanics and nuclear design and discusses a few of the specific structural mechanics problems faced by the nuclear designers in the Department of Atomic Energy, India. (author)

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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.)

  1. 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.

  2. 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.

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. Nonlinear effects in dynamic analysis and design of nuclear power plant components: research status and needs

    Energy Technology Data Exchange (ETDEWEB)

    Stoykovich, M [Burns and Roe, Inc., New York (USA)

    1978-10-01

    This paper encompasses nonlinear effects in dynamic analysis and design of nuclear power plant facilities. The history of plasticity as a science is briefly discussed, and nonlinear cases of special interest are described. Approaches to some of the nonlinear problems are presented. These include the nonlinearity due to foundation-structure interaction associated with the base slab uplift during seismic disturbances, the nonlinear base-isolation system for the reduction of earthquake-generated forces and deformations of superstructures, nonlinear systems having restoring-force functions in case of gaps and liift-off conditions, and nonlinearity of viscoelastic systems due to inelastic deformations. Available computer programs information for the solution of various types of nonlinear problems are provided. Advantages and disadvantages of some of the nonlinear and linear analyses are discussed. Comparison of some nonlinear and linear results of analyses are presented. Conclusions are reached with regard to research status and recommendations for further studies and for performing non-linear analyses associated with the problems of nonlinearity are presented.

  8. Nonlinear effects in dynamic analysis and design of nuclear power plant components: research status and needs

    International Nuclear Information System (INIS)

    Stoykovich, M.

    1978-01-01

    This paper encompasses nonlinear effects in dynamic analysis and design of nuclear power plant facilities. The history of plasticity as a science is briefly discussed, and nonlinear cases of special interest are described. Approaches to some of the nonlinear problems are presented. These include the nonlinearity due to foundation-structure interaction associated with the base slab uplift during seismic disturbances, the nonlinear base-isolation system for the reduction of earthquake-generated forces and deformations of superstructures, nonlinear systems having restoring-force functions in case of gaps and liift-off conditions, and nonlinearity of viscoelastic systems due to inelastic deformations. Available computer programs information for the solution of various types of nonlinear problems are provided. Advantages and disadvantages of some of the nonlinear and linear analyses are discussed. Comparison of some nonlinear and linear results of analyses are presented. Conclusions are reached with regard to research status and recommendations for further studies and for performing non-linear analyses associated with the problems of nonlinearity are presented. (Auth.)

  9. Nuclear analysis

    International Nuclear Information System (INIS)

    1988-01-01

    In a search for correlations between the elemental composition of trace elements in human stones and the stone types with relation to their growth pattern, a combined PIXE and x-ray diffraction spectrometry approach was implemented. The combination of scanning PIXE and XRD has proved to be an advance in the methodology of stone analysis and may point to the growth pattern in the body. The exact role of trace elements in the formation and growth of urinary stones is not fully understood. Efforts are thus continuing firstly to solve the analytical problems concerned and secondly to design suitable experiments that would provide information about the occurrence and distribution of trace elements in urine. 1 fig., 1 ref

  10. 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

  11. 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

  12. Human factors analysis and design methods for nuclear waste retrieval systems. Volume II. A compendium of human factors design data

    International Nuclear Information System (INIS)

    Casey, S.M.

    1980-04-01

    This document is a compilation of human factors engineering design recommendations and data, selected and organized to assist in the design of a nuclear waste retrieval system. Design guidelines from a variety of sources have been evaluated, edited, and expanded for inclusion in this document, and, where appropriate, portions of text from selected sources have been included in their entirety. A number of human factors engineering guidelines for equipment designers have been written over the past three decades, each tailored to the needs of the specific system being designed. In the case of this particular document, a review of the preliminary human operator functions involved in each phase of the retrieval process was performed, resulting in the identification of areas of design emphasis upon which this document should be based. Documents containing information and design data on each of these areas were acquired, and data and design guidelines related to the previously identified areas of emphasis were extracted and reorganized. For each system function, actions were first assigned to operator and/or machine, and the operator functions were then described. Separate lists of operator functions were developed for each of the areas of retrieval activities - survey and mapping, remining, floor flange emplacement, plug and canister overcoring, plug and canister removal and transport, and CWSRS activity. These functions and the associated man-machine interface were grouped into categories based on task similarity, and the principal topics of human factors design emphasis were extracted. These topic areas are reflected in the contents of the 12 sections of this document

  13. Design and evaluation of augmented reality for nuclear power plant maintenance based on Cognitive activity analysis

    International Nuclear Information System (INIS)

    Koo, Jwa Jin

    2008-02-01

    The goals of maintenance in nuclear power plants (NPPs) are to reduce the number of plant shutdowns, to decrease the shutdown time, to minimize the maintenance costs, and to improve nuclear safety. Furthermore, it is difficult for maintenance personnel to conduct work in NPPs due to the complexity of NPPs, safety issues and time constraints. Human errors should be prevented to achieve maintenance goals in NPPs. Therefore, it is necessary to improve working conditions by introducing support systems. Augmented Reality (AR) has the potential to assist with maintenance tasks and make maintenance personnel concentrate on the tasks, which will then the reduced human error and improve human performance. The working procedure for AR systems was designed based on the characteristics of cognitive activity, and an experiment was conducted to estimate the effectiveness of AR. The results indicated that a well-designed working procedure for AR systems was more effective than a conventional paper-based working procedure. In addition, participants made fewer errors using the designed working procedure. AR systems as maintenance support systems are beneficial technology that can improve maintenance performance by reducing cognitive loads and unnecessary movements. An AR application can be attractive and effective at supporting maintenance task

  14. 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.)

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Lerner, A.M.

    1996-01-01

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  1. 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

  2. General aspects of design and vessel nozzle analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Back, N.

    1980-01-01

    Aspects of design and a procedure for nozzle tensile analysis under loads in project, normal and abnormal, emergency, failure and test conditions. For each condition, considerations about the tensile calculation methods, the tensile classification in corresponding categories and the comparison with admissible limits according to the norms. (M.C.K.) [pt

  3. Safety Effect Analysis of the Large-Scale Design Changes in a Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eun-Chan; Lee, Hyun-Gyo [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    These activities were predominantly focused on replacing obsolete systems with new systems, and these efforts were not only to prolong the plant life, but also to guarantee the safe operation of the units. This review demonstrates the safety effect evaluation using the probabilistic safety assessment (PSA) of the design changes, system improvements, and Fukushima accident action items for Kori unit 1 (K1). For the large scale of system design changes for K1, the safety effects from the PSA perspective were reviewed using the risk quantification results before and after the system improvements. This evaluation considered the seven significant design changes including the replacement of the control building air conditioning system and the performance improvement of the containment sump using a new filtering system as well as above five system design changes. The analysis results demonstrated that the CDF was reduced by 12% overall from 1.62E-5/y to 1.43E-5/y. The CDF reduction was larger in the transient group than in the loss of coolant accident (LOCA) group. In conclusion, the analysis using the K1 PSA model supports that the plant safety has been appropriately maintained after the large-scale design changes in consideration of the changed operation factors and failure modes due to the system improvements.

  4. 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

  5. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    International Nuclear Information System (INIS)

    Labib, Satira; King, Jeffrey

    2015-01-01

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort

  6. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  7. 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.

  8. 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.

  9. 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)

  10. Design and analysis of aquatic monitoring programs at nuclear power plants

    International Nuclear Information System (INIS)

    McKenzie, D.H.; Kannberg, L.D.; Gore, K.L.; Arnold, E.M.; Watson, D.G.

    1977-11-01

    This report addresses some of the problems of designing, conducting, and analyzing aquatic environmental monitoring programs for impact assessment of nuclear power plants. The concepts discussed are applicable to monitoring the effects of chemical, radioactive, or thermal effluents. The concept of control and treatment station pairs is the fundamental basis for the experimental method proposed. This concept is based on the hypothesis that the relationship between the two stations forming the pair can be estimated from the preoperational period and that this relationship holds during the operational period. Any changes observed in this relationship during the operational period are assumed to be the result of the power plant impacts. Thus, it is important that station pairs are selected so it can be assumed that they respond to natural environmental changes in a manner that maintains that relationship. The major problem in establishing the station pairs will be the location of the control station. The universal heterogeneity in the environment will prevent the establishment of identical station pairs. The requirement that the control station remain unaffected by the operation of the power plant dictates a spacial separation with its associated differences in habitat. Thus, selection of the control station will be based upon balancing the following two criteria: (1) far enough away from the plant site to be beyond the plant influence, and (2) close enough to the treatment station that the biological communities will respond to natural environmental changes consistently in the same manner

  11. Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

    2002-03-01

    The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

  12. 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.

  13. Improvement of reference nuclear data for commercial power reactor analysis and design. Final report

    International Nuclear Information System (INIS)

    Levitt, L.; Magurno, B.; Rose, P.

    1977-10-01

    The National Neutron Cross Section Center (NNCSC) has participated in the Electric Power Research Institute (EPRI) program of improving the nuclear data base to be used as input for the analysis of power reactors. The NNCSC contribution consisted of organizing and participating in specialists' meetings on the Thermal Neutron Cross Sections for the Actinides and Fission Products, participating in a meeting on the recommended value for the 252 Cf anti nu, and keeping abreast of new measurements on the low energy resonance parameters of 238 U. In addition, complete evaluations for seven isotopes of gadolinium have been produced for the next version of the data library. Definitive benchmark calculations using the Monte Carlo code RECAP have been carried out for two homogeneous uranium nitrate critical spheres (ORNL 1 and 2), two water moderated uranium metal lattices (TRX-1 and 2), and three plutonium nitrate spheres (PNL 2, 3, and 5). Results for the uranium spheres gave good agreement with experiments and with previous calculations. The calculation of the TRX lattice parameters, however, casts doubt on the validity of analytic leakage corrections used previously. Eigenvalues of the plutonium nitrate spheres are found to increase with increasing ratio of total 240 Pu/ 239 Pu captures

  14. Design base transient analysis using the real-time nuclear reactor simulator model

    International Nuclear Information System (INIS)

    Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

    1987-01-01

    A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and moderator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

  15. Design base transient analysis using the real-time nuclear reactor simulator model

    International Nuclear Information System (INIS)

    Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

    1987-01-01

    A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and mode-rator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

  16. A compact neutron beam generator system designed for prompt gamma nuclear activation analysis.

    Science.gov (United States)

    Ghassoun, J; Mostacci, D

    2011-08-01

    In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5). 2010 Elsevier Ltd. All rights reserved.

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. Proceedings of the Conference on Structural Analysis and Design of Nuclear Power Plants. v. 2

    International Nuclear Information System (INIS)

    1984-01-01

    It's presented the discussion of recents research results and developments related to applications of structural mechanics and seismology to mechanical structures and to nuclear facility sites. (E.G.) [pt

  2. Proceedings of the Conference on Structural Analysis and Design of Nuclear Power Plants. v. 1

    International Nuclear Information System (INIS)

    1984-01-01

    It's presented the discussion of recents research results and developments related to applications of structural mechanics to nuclear power plants, in particular, and to conventional power plants, in a more general context. (E.G.) [pt

  3. Analysis of design of auxiliary system of Booshehr Nuclear Power Plant

    International Nuclear Information System (INIS)

    Naseh Hasanzadeh, M.

    1999-01-01

    Power plant's internal auxiliary system has an important role in its safety operation. Because of the decay heat and safety aspects in the nuclear power plants, this role is more important. In this thesis, operation of the nuclear power plant with PWR reactor is studied and deferent nuclear systems described. In the next section all electrical loads in the Booshehr Nuclear Power Plant identified and feeding methods of each load is determined. by use of the single line diagram of the internal auxiliary system, the nominal rating of all electrical devices as transformers, inverters, Ups, diesel generators and etc. is determined. In the following, short circuit calculations performed and by above conclusion, rating values of circuit breakers is determined. At last the starting problems of electrical motors is studied and the results of motor's behavior at starting moment is discussed

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. AZTLAN: Mexican platform for analysis and design of nuclear reactors - 15493

    International Nuclear Information System (INIS)

    Gomez Torres, A.M.; Puente Espel, F.; Valle Gallegos, E. del; Francois, J.L.; Martin-del-Campo, C.; Espinosa-Paredes, G.

    2015-01-01

    The AZTLAN platform is presented in this paper. This project aims at modernizing, improving and incorporating the neutron transport codes such as AZTLAN, AZKIND and AZNHEX, thermo-hydraulics codes like AZTHECA and thermo-mechanical codes developed in the Mexican institutions of higher education as well as in the Mexican nuclear research institute, in an integrated platform, established and maintained for the benefit of the Mexican nuclear knowledge. An important part of the project is to develop a coupling methodology between neutron transport codes and thermal-hydraulics codes in order to get an accurate 3-dimensional simulation of a reactor core

  10. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  11. 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)

  12. 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

  13. Design and analysis of free-standing spent fuel racks in nuclear power plants

    International Nuclear Information System (INIS)

    Ashar, H.; DeGrassi, G.

    1989-01-01

    With the prohibition on reprocessing of spent fuel in the late 1970's the pools which were supposed to be short term storage became quasi-permanent storage spaces for spent fuel. Recognizing a need to provide permanent storage facilities for such nuclear wastes, the US Congress enacted a law cited as the Nuclear Waste Policy Act of 1982. The Act, in essence, required the Department of Energy to find ways for long term storage of high level waste. However, it also is required the owners of nuclear power plants to provide for interim storage of their spent fuel. The permanent government owned repositories are not scheduled to be operational until the year 2005. In order to accommodate the increasing inventory of spent fuel, the US utilities started looking for various means to store spent fuel at the reactor sites. One of the most economical ways to accommodate more spent fuel is to arrange storage locations as closely as possible at the same time making sure that the fuel remains subcritical and that there are adequate means to cope with the heat load. The free standing high density rack configuration is an outcome of efforts to accommodate to more fuel in the limited space. 3 refs., 3 figs

  14. 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

  15. 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.

  16. Deuteron nuclear data for the design of accelerator-based neutron sources: Measurement, model analysis, evaluation, and application

    Science.gov (United States)

    Watanabe, Yukinobu; Kin, Tadahiro; Araki, Shouhei; Nakayama, Shinsuke; Iwamoto, Osamu

    2017-09-01

    A comprehensive research program on deuteron nuclear data motivated by development of accelerator-based neutron sources is being executed. It is composed of measurements of neutron and gamma-ray yields and production cross sections, modelling of deuteron-induced reactions and code development, nuclear data evaluation and benchmark test, and its application to medical radioisotopes production. The goal of this program is to develop a state-of-the-art deuteron nuclear data library up to 200 MeV which will be useful for the design of future (d,xn) neutron sources. The current status and future plan are reviewed.

  17. Design Analysis of a Prepackaged Nuclear Power Plant for an Ice Cap Location

    Science.gov (United States)

    1959-01-15

    to any rod drive component. Alco designed and built for ANL a prototype drive for the ALPR , This t prototype was extensively tested and approved...the Alco built ALPR rod drives now in operation at Arco, Rod position indication is by means of 2 synchros which are gear driven on the rod side

  18. 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)

  19. 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)

  20. Conceptual design of an integrated information system for safety related analysis of nuclear power plants (IRIS Phase 1)

    International Nuclear Information System (INIS)

    Hofer, K.; Zehnder, P.; Galperin, A.

    1994-01-01

    This report deals with a conceptual design of an integrated information management system, called PSI-IRIS, as needed to assist the analysts for reactor safety related investigations on Swiss nuclear power plants within the project STARS. Performing complicated engineering analyses of an NPP requires storage and manipulation of a large amount of information, both data and knowledge. This information is characterized by its multi-disciplinary nature, complexity, and diversity. The problems caused by inefficient and lengthy manual operations involving the data flow management within the framework of the safety related analysis of an NPP, can be solved by applying computer aided engineering (CAE) principles. These principles are the basis for the design of the integrated information management system PSI-IRIS presented in this report. The basic idea is to create a computerized environment, which includes both database and functional capabilities. The database of the PSI-IRIS consists of two parts, an NPP generic database (GDB) and a collection of analysis results (CASE L IB). The GDB includes all technical plant data and information needed to generate input decks for all computer codes utilized within the STARS project. The CASE L IB storage contains the accumulated knowledge, input decks, and result files of the NPP transient analyses. Considerations and analysis of the data types and the required data manipulation capabilities as well as operational requirements resulted in the choice of an object-oriented database management system (OODBMS) as a development platform for solving the software engineering problems. Several advantages of OODBMS's over conventional relational database management systems were found of crucial importance, especially providing the necessary flexibility for different data types and the potential for extensibility. (author) 15 figs., tabs., 20 refs

  1. Structural analysis and design of a nuclear power plant building for aircraft crash effects

    International Nuclear Information System (INIS)

    Degen, P.; Furrer, H.; Jemielewski, J.

    1976-01-01

    The object of this investigation is to assess the effect of a large commercial airplane crashing perpendicularly on to the surface of a spherical reactor building dome. This investigation is related to a project currently in execution. Practical solutions of the postulated case, which vary in the degree of engineering effort used, are shown. Based on safety consideration the various solutions are discussed from the viewpoint of penetration, cracking and collapse modes of failure, where, primarily, the carrying capacity of the structure under an equivalent statical load is considered. The performed investigations include: (a) Calculation of the failure load following the yield line theory; (b) Calculation of the sectional forces using the linear-elastic shell theory and subsequent design by the ultimate strength method; (c) Calculation of the failure load, establishing of the failure mechanism and distribution of sectional forces using the plastic shell theory; (d) Calculation using a three-dimensional FEM program with plastic capabilities; this includes the collapse load, the failure mechanism and the distribution of sectional forces. A discussion of the resultant forces and the configuration of the critical section is given for the various methods used. The evaluation of the carrying capacity of the structure with respect to load is based on energy considerations. It is attempted to compare such results, to evaluate possible simplifications in the used solutions, and to give some recommendations for the practical design and for the development of structural details. (Auth.)

  2. 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

  3. 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.)

  4. 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

  5. Highlights in KMRR nuclear design

    International Nuclear Information System (INIS)

    Kim, H.R.; Kim, B.G.; Kim, Y.J.; Lee, B.C.; Lee, J.T.; Lee, J.B.

    1990-01-01

    The Korea Advanced Energy Research Institute (KAERI) is constructing the KMRR (Korea Multi-purpose Research Reactor) to meet the requirements of the nationwide nuclear programs. The KMRR is an open-tank-in-pool type reactor which uses LEU U 3 Si-Al fuel and features an LWR/HWR hybrid configuration that has advantage of easily tailoring experimental facility layout to ensure supply of effective neutron quality. KAERI established the physics design system including the shielding design specifically for the KMRR. Based on the design analyse the reactor can be operated at a thermal power output of 30 MW to suit applications ranging from material/fuel testing to beam utilization, without violating the design limits concerning fuel integrity, and the radioactivity level in the access area is kept below the design criteria of 1.25 mrem/hr

  6. Design, analysis and construction of the prestressed concrete containment of the nuclear power station Gundremmingen

    International Nuclear Information System (INIS)

    Mueller, W.F.; Ick, U.

    1977-01-01

    Kraftwerk Union AG is presently constructing at Gundremmingen (Bavaria) on the River Danube a BWR twin-plant (KRB Units B and C) with a capacity of 2x1300 MWe. Owing to the wall thickness/diameter ratio the containment can be calculated as a thin-walled shell. Areas of discontinuity are subjected to three-dimensional investigations. For the design of the concrete structure different fracture safety margins are defined for the load conditions occurring in operation in the event of a loss-of-coolant accident and as a result of an aircraft or an earthquake. From this results that in the cross sectional areas without discontinuities of the prestressed outer cylinder no resultant tensions occur. For the steel liner different limits of strain are permitted for the various load conditions, bearing in mind that the integrity of the liner must remain ensured at any time. In order to keep the stresses resulting from the constraint of the containment outer cylinder in the foundation slab low, the cylindrical wall is placed on bearings. The suppression pool top slab is constrained at the containment outer cylinder and at the containment inner cylindrical wall. The inlets of the vent pipes are integrated in the slab in a way resulting in a double slab. The liner consists of 8 mm thick steel plate and is anchored in the concrete via steel sections. Mechanical equipment anchoring in the concrete is provided by welding anchor plates into the liner after the section concerned has been completed. The carcass work on the reactor building is scheduled to be completed within

  7. 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

  8. IEEE standard requirements for reliability analysis in the design and operation of safety systems for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The purpose of this standard is to provide uniform, minimum acceptable requirements for the performance of reliability analyses for safety-related systems found in nuclear-power generating stations, but not to define the need for an analysis. The need for reliability analysis has been identified in other standards which expand the requirements of regulations (e.g., IEEE Std 379-1972 (ANSI N41.2-1972), ''Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection System,'' which describes the application of the single-failure criterion). IEEE Std 352-1975, ''Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems,'' provides guidance in the application and use of reliability techniques referred to in this standard

  9. Design of nuclear power plants

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1987-01-01

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

  10. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  11. 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)

  12. 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

  13. 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

  14. 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

  15. CFD analysis and experimental investigation associated with the design of the Los Alamos nuclear materials storage facility

    International Nuclear Information System (INIS)

    Bernardin, J.D.; Hopkins, S.; Gregory, W.S.; Martin, R.A.

    1997-01-01

    The Nuclear Materials Storage Facility (NMSF) at the Los Alamos National Laboratory is being renovated for long-term storage of canisters designed to hold heat-generating nuclear materials, such as powders, ingots, and other components. The continual heat generation within the canisters necessitates a reliable cooling scheme of sufficient magnitude which maintains the stored material temperatures within acceptable limits. The primary goal of this study was to develop both an experimental facility and a computational fluid dynamics (CFD) model of a subsection of the NMSF which could be used to observe general performance trends of a proposed passive cooling scheme and serve as a design tool for canister holding fixtures. Comparisons of numerical temperature and velocity predictions with empirical data indicate that the CFD model provides an accurate representation of the NMSF experimental facility. Minor modifications in the model geometry and boundary conditions are needed to enhance its accuracy, however, the various fluid and thermal models correctly capture the basic physics

  16. Equivalent linear and nonlinear site response analysis for design and risk assessment of safety-related nuclear structures

    International Nuclear Information System (INIS)

    Bolisetti, Chandrakanth; Whittaker, Andrew S.; Mason, H. Benjamin; Almufti, Ibrahim; Willford, Michael

    2014-01-01

    Highlights: • Performed equivalent linear and nonlinear site response analyses using industry-standard numerical programs. • Considered a wide range of sites and input ground motions. • Noted the practical issues encountered while using these programs. • Examined differences between the responses calculated from different programs. • Results of biaxial and uniaxial analyses are compared. - Abstract: Site response analysis is a precursor to soil-structure interaction analysis, which is an essential component in the seismic analysis of safety-related nuclear structures. Output from site response analysis provides input to soil-structure interaction analysis. Current practice in calculating site response for safety-related nuclear applications mainly involves the equivalent linear method in the frequency-domain. Nonlinear time-domain methods are used by some for the assessment of buildings, bridges and petrochemical facilities. Several commercial programs have been developed for site response analysis but none of them have been formally validated for large strains and high frequencies, which are crucial for the performance assessment of safety-related nuclear structures. This study sheds light on the applicability of some industry-standard equivalent linear (SHAKE) and nonlinear (DEEPSOIL and LS-DYNA) programs across a broad range of frequencies, earthquake shaking intensities, and sites ranging from stiff sand to hard rock, all with a focus on application to safety-related nuclear structures. Results show that the equivalent linear method is unable to reproduce the high frequency acceleration response, resulting in almost constant spectral accelerations in the short period range. Analysis using LS-DYNA occasionally results in some unrealistic high frequency acceleration ‘noise’, which can be removed by smoothing the piece-wise linear backbone curve. Analysis using DEEPSOIL results in abrupt variations in the peak strains of consecutive soil layers

  17. Contribution of the ergonomic analysis to the improvement of the design of operating procedures in nuclear power plants

    International Nuclear Information System (INIS)

    Dien, Y.; Montmayeul, R.

    1992-11-01

    The design of operating procedures for continuous processes is much too often based on implicit assumptions both concerning the operators and the operating conditions that must be dealt with. The merit of the ergonomic approach to the design of procedures is to take account of the way the various operators actually use operating procedures. The actual use is determined from the analysis of on-site operation (normal and incident operating conditions) and the analysis of full-scale simulators tests (incident operating conditions). The introduction of the ergonomic approach in the procedure design results in new design principles being proposed

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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.

  3. 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

  4. Economic Analysis of Nuclear Energy

    International Nuclear Information System (INIS)

    Lee, Man Ki; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Oh, K. B.

    2006-12-01

    It has been well recognized that securing economic viabilities along with technologies are very important elements in the successful implementation of nuclear R and D projects. The objective of the Project is to help nuclear energy to be utilized in an efficient way by analyzing major issues related with nuclear economics. The study covers following subjects: the role of nuclear in the future electric supply system, economic analysis of nuclear R and D project, contribution to the regional economy from nuclear power. In addition, the study introduces the international cooperation in the methodological area of efficient use of nuclear energy by surveying the international activities related with nuclear economics

  5. Economic Analysis of Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Man Ki; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Oh, K. B

    2006-12-15

    It has been well recognized that securing economic viabilities along with technologies are very important elements in the successful implementation of nuclear R and D projects. The objective of the Project is to help nuclear energy to be utilized in an efficient way by analyzing major issues related with nuclear economics. The study covers following subjects: the role of nuclear in the future electric supply system, economic analysis of nuclear R and D project, contribution to the regional economy from nuclear power. In addition, the study introduces the international cooperation in the methodological area of efficient use of nuclear energy by surveying the international activities related with nuclear economics.

  6. 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.

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. Nuclear power regional analysis

    International Nuclear Information System (INIS)

    Parera, María Delia

    2011-01-01

    In this study, a regional analysis of the Argentine electricity market was carried out considering the effects of regional cooperation, national and international interconnections; additionally, the possibilities of insertion of new nuclear power plants in different regions were evaluated, indicating the most suitable areas for these facilities to increase the penetration of nuclear energy in national energy matrix. The interconnection of electricity markets and natural gas due to the linkage between both energy forms was also studied. With this purpose, MESSAGE program was used (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts), promoted by the International Atomic Energy Agency (IAEA). This model performs a country-level economic optimization, resulting in the minimum cost for the modelling system. Regionalization executed by the Wholesale Electricity Market Management Company (CAMMESA, by its Spanish acronym) that divides the country into eight regions. The characteristics and the needs of each region, their respective demands and supplies of electricity and natural gas, as well as existing and planned interconnections, consisting of power lines and pipelines were taken into account. According to the results obtained through the model, nuclear is a competitive option. (author) [es

  15. 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.

  16. 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.)

  17. Innovative designs of nuclear reactors

    International Nuclear Information System (INIS)

    Gabaraev, B.A.; Cherepnin, Y.S.

    2010-01-01

    The world development scenarios predict at least a 2.5 time increase in the global consumption of primary energy in the first half of the twenty-first century. Much of this growth can be provided by the nuclear power which possesses important advantages over other energy technologies. However, the large deployment of nuclear sources may take place only when the new generation of reactors appears on the market and will be free of the shortcomings found in the existing nuclear power installations. The public will be more inclined to accept nuclear plants that have better economics; higher safety; more efficient management of the radioactive waste; lower risk of nuclear weapons proliferation, and provided that the focus is made on the energy option free of ∇ e 2 generation. Currently, the future of nuclear power is trusted to the technology based on fast reactors and closed fuel cycle. The latter implies reprocessing of the spent nuclear fuel of the nuclear plants and re-use of plutonium produced in power reactors

  18. 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)

  19. 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

  20. The design of the frame structure used in integral hosting of the nuclear island steel lining cylinder module and problems analysis

    International Nuclear Information System (INIS)

    Yu Xinian; Liu Xiao; Wang Jianguo

    2011-01-01

    The use of the steel frame in the integral hosting of nuclear island steel lining cylinder module made a breakthrough in China's nuclear power construction. The deformation of the cylinder wall is the key issue in the integral lifting process of the nuclear island steel lining. Using the frame in lifting large and thin steel cylinder, the form of frame structure and its deformation will directly affect the radial deformation of the lifted cylinder, the buckling deformation of the distal cylinder, and the cylinder's deformation surround the penetrations. The diameter of nuclear island steel liner is 44 meters. The wall of the cylinder is thin, and the total weight of the cylinder itself and its attached penetrations, walkways and lifting tools, etc. is up to 120 tons, which not only increase the difficulty of lifting, but also have some risks. To ensure the cylinder deformation within the limits, this thesis establishes the parameter structure for the lifting frame, calculates the displacement and analyzes the axial stresses, based on the ANSYS finite element analysis software. The results showed that the models and parameters for integral hosting of the steel lining cylinder modular frame structure is reasonable and feasible, and analyzing the hosting-frame data is necessary, which lay the foundation for the design of the hosting frame and the eventual implementation of the integral hosting scheme of the steel lining cylinder module. (authors)

  1. 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.

  2. 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)

  3. 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.)

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  9. Seismic-design questions typify nuclear obstacles

    International Nuclear Information System (INIS)

    Strauss, S.D.

    1979-01-01

    The trade-off between safe design of nuclear power plants and cost is considered. As an example, seismic protection problems at the Beaver Valley station of Duquesne Light Co. and their resolution by Stone and Webster Engineering are discussed

  10. 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

  11. 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)

  12. Advances in the development of the Mexican platform for analysis and design of nuclear reactors: AZTLAN Platform; Avances en el desarrollo de la plataforma mexicana para analisis y diseno de reactores nucleares: AZTLAN Platform

    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, 07738 Ciudad de Mexico (Mexico); Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: armando.gomez@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2017-09-15

    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)

  13. System analysis and design

    International Nuclear Information System (INIS)

    Son, Seung Hui

    2004-02-01

    This book deals with information technology and business process, information system architecture, methods of system development, plan on system development like problem analysis and feasibility analysis, cases for system development, comprehension of analysis of users demands, analysis of users demands using traditional analysis, users demands analysis using integrated information system architecture, system design using integrated information system architecture, system implementation, and system maintenance.

  14. 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

  15. 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

  16. IRIS Nuclear Power Plant design

    International Nuclear Information System (INIS)

    Carelli, M. D.; Cobian, J.

    2002-01-01

    IRIS(International Reactor Innovative and Secure) is a novel light water reactor with a modular, integral primary system configuration. This concept, initially developed in response to the first NERI solicitation, is now being pursued by an international consortium of 20 participants from seven countries. IRIS is designed to satisfy the four key requirements for Generation IV systems: enhanced safety, improved economics, proliferation resistance and waste minimization. Its main features are: small-to-medium power (100-335 MWe/module); long life core 5 to 10 years) without shuffling or refueling; optimized maintenance with repair shutdown intervals of a least four years; simplified compact design with the primary vessel housing steam generators, pressurizer and pumps; safety by design where accidents are positively eliminated by design rather than engineering to cope with their consequences; loss of coolant accidents of any size and loss of low accidents are eliminated as major safety concerns; estimated power generation total cost is projected to be competitive with other power options. IRIS is one of four new reactor designs currently under NRC review. Projected schedule calls for design certification by 2008 and being ready for deployment by 2001 or later. This rather short schedule is made possible by the fact that IRIS is based on proven light water technology and new technology development is not required. (Author)

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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.)

  3. Function Allocation in Complex Socio-Technical Systems: Procedure usage in nuclear power and the Context Analysis Method for Identifying Design Solutions (CAMIDS) Model

    Science.gov (United States)

    Schmitt, Kara Anne

    This research aims to prove that strict adherence to procedures and rigid compliance to process in the US Nuclear Industry may not prevent incidents or increase safety. According to the Institute of Nuclear Power Operations, the nuclear power industry has seen a recent rise in events, and this research claims that a contributing factor to this rise is organizational, cultural, and based on peoples overreliance on procedures and policy. Understanding the proper balance of function allocation, automation and human decision-making is imperative to creating a nuclear power plant that is safe, efficient, and reliable. This research claims that new generations of operators are less engaged and thinking because they have been instructed to follow procedures to a fault. According to operators, they were once to know the plant and its interrelations, but organizationally more importance is now put on following procedure and policy. Literature reviews were performed, experts were questioned, and a model for context analysis was developed. The Context Analysis Method for Identifying Design Solutions (CAMIDS) Model was created, verified and validated through both peer review and application in real world scenarios in active nuclear power plant simulators. These experiments supported the claim that strict adherence and rigid compliance to procedures may not increase safety by studying the industry's propensity for following incorrect procedures, and when it directly affects the outcome of safety or security of the plant. The findings of this research indicate that the younger generations of operators rely highly on procedures, and the organizational pressures of required compliance to procedures may lead to incidents within the plant because operators feel pressured into following the rules and policy above performing the correct actions in a timely manner. The findings support computer based procedures, efficient alarm systems, and skill of the craft matrices. The solution to

  4. 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

  5. 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

  6. 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

  7. 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)

  8. Nuclear power ecology: comparative analysis

    International Nuclear Information System (INIS)

    Trofimenko, A.P.; Lips'ka, A.Yi.; Pisanko, Zh.Yi.

    2005-01-01

    Ecological effects of different energy sources are compared. Main actions for further nuclear power development - safety increase and waste management, are noted. Reasons of restrained public position to nuclear power and role of social and political factors in it are analyzed. An attempt is undertaken to separate real difficulties of nuclear power from imaginary ones that appear in some mass media. International actions of environment protection are noted. Risk factors at different energy source using are compared. The results of analysis indicate that ecological influence and risk for nuclear power are of minimum

  9. Analysis of availability, functional integration and remote maintenance for the design of critical components and systems in nuclear fusion technology

    OpenAIRE

    Fernández Berceruelo, Iván

    2016-01-01

    Junto a las conocidas ventajas de la Fusión Nuclear frente a otras tecnologías energéticas: seguridad inherente por principio físico, gestión del impacto ambiental y disponibilidad a largo plazo de combustibles primarios (deuterio, litio), la competitividad de la energía generada por fusión determinará su capacidad de penetración en el mercado. Para lograrla será necesario revisar rigurosamente las opciones de diseño existentes, de modo que se justifique sólidamente la elección de una solució...

  10. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

  11. 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

  12. Economic Analysis of Nuclear Energy

    International Nuclear Information System (INIS)

    Kim, S. S.; Lee, M. K.; Moon, K. H.; Nam, J. H.; Noh, B. C.; Kim, H. R.

    2008-12-01

    The concerns on the global warming issues in the international community are bringing about a paradigm shift in the national economy including energy technology development. In this connection, the green growth mainly utilizing green technology, which emits low carbon, is being initiated by many advanced countries including Korea. The objective of the study is to evaluate the contribution to the national economy from nuclear energy attributable to the characteristics of green technology, to which nuclear energy belongs. The study covers the role of nuclear in addressing climate change issues, the proper share of nuclear in the electricity sector, the cost analyses of decommissioning and radioactive waste management, and the analysis on the economic performance of nuclear R and D including cost benefit analysis

  13. 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)

  14. 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

  15. 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.

  16. 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)

  17. 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)

  18. Nuclear design of APSARA reload-2 fuel

    International Nuclear Information System (INIS)

    Nath, M.; Veeraraghavan, N.

    1978-01-01

    In view of the satisfactory operating performance of initial and reload-1 fuel designs of Apsara reactor, it was felt desirable to adopt a basically similar design for reload-2 fuel, i.e. the fuel assembly should consist of equally spaced parallel fuel plates in which highly enriched uranium, alloyed with aluminium, is employed as fuel. However, because of fabricational constraints, certain modifications were necessary and were incorporated in the proposed reload design to cater to the multiple needs of operational requirements, improved fuel utilization and inherent reactor safety. The salient features of the nuclear design of reload-2 fuel for the Apsara reactor are discussed. (author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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.)

  6. 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.

  7. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    OpenAIRE

    Genard Gilles; Portal Romain; Bouchat Virginie; Vanderperre Serge

    2017-01-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assump...

  8. 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

  9. 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

  10. 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

  11. Design and Analysis Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides engineering design of aircraft components, subsystems and installations using Pro/E, Anvil 1000, CADKEY 97, AutoCAD 13. Engineering analysis tools include...

  12. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Science.gov (United States)

    Genard, Gilles; Portal, Romain; Bouchat, Virginie; Vanderperre, Serge

    2017-09-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  13. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Genard Gilles

    2017-01-01

    Full Text Available In the framework of the design studies for a new Emergency Operation Facility (EOF for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  14. 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.)

  15. Nuclear forensic analysis of thorium

    International Nuclear Information System (INIS)

    Moody, K.J.; Grant, P.M.

    1999-01-01

    A comprehensive radiochemical isolation procedure and data analysis/interpretation method for the nuclear forensic investigation of Th has been developed. The protocol includes sample dissolution, chemical separation, nuclear counting techniques, consideration of isotopic parent-daughter equilibria, and data interpretation tactics. Practical application of the technology was demonstrated by analyses of a questioned specimen confiscated at an illegal drug synthesis laboratory by law enforcement authorities. (author)

  16. Multidimensional Analysis of Nuclear Detonations

    Science.gov (United States)

    2015-09-17

    Training Command in Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy Robert C. Slaughter, B.S., M.S. Captain, USAF 16...15-S-029 Abstract Digitized versions of atmospheric nuclear testing films represent a unique data set that enables the scientific community to create...temperature distribution of a nuclear fireball using digitized film . This temperature analysis underwent verification using the Digital Imaging and Remote

  17. 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

  18. 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)

  19. 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)

  20. Human factors analysis and design methods for nuclear waste retrieval systems. Volume III. User's guide for the computerized event-tree analysis technique

    International Nuclear Information System (INIS)

    Casey, S.M.; Deretsky, Z.

    1980-08-01

    This document provides detailed instructions for using the Computerized Event-Tree Analysis Technique (CETAT), a program designed to assist a human factors analyst in predicting event probabilities in complex man-machine configurations found in waste retrieval systems. The instructions contained herein describe how to (a) identify the scope of a CETAT analysis, (b) develop operator performance data, (c) enter an event-tree structure, (d) modify a data base, and (e) analyze event paths and man-machine system configurations. Designed to serve as a tool for developing, organizing, and analyzing operator-initiated event probabilities, CETAT simplifies the tasks of the experienced systems analyst by organizing large amounts of data and performing cumbersome and time consuming arithmetic calculations. The principal uses of CETAT in the waste retrieval development project will be to develop models of system reliability and evaluate alternative equipment designs and operator tasks. As with any automated technique, however, the value of the output will be a function of the knowledge and skill of the analyst using the program

  1. Seismic analysis of design

    International Nuclear Information System (INIS)

    Jehlicka, P.

    1980-01-01

    The determination of the dynamic response of nuclear power plants is a necessary part of safe design against earthquake, or against other additional vibrational loading. The determination of these dynamic loads caused by external excitation is a requirement in calculating the related material loading on the structures. The purpose of this lecture is to present a general survey of analytical methods to determine the response of structural and mechanical equipment to earthquake. The main problems which complicate structural-dynamic calculations will be discussed. The necessity to control input parameters and the possibility to calculate with simplified methods will be pointed out. (orig./RW)

  2. 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

  3. 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

  4. Economic analysis of nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S

    2000-12-01

    This study identified the role of nuclear energy in the following three major aspects. First of all, this study carried out cost effectiveness of nuclear as a CDM technology, which is one of means of GHG emission reduction in UNFCCC. Secondly, environmental externalities caused by air pollutants emitted by power options were estimated. The 'observed market behaviour' method and 'responses to hypothetical market' method were used to estimate objectively the environmental external costs by electric source, respectively. Finally, the role of nuclear power in securing electricity supply in a liberalized electricity market was analyzed. This study made efforts to investigate whether nuclear power generation with high investment cost could be favored in a liberalized market by using 'option value' analysis of investments.

  5. Economic analysis of nuclear energy

    International Nuclear Information System (INIS)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S.

    2000-12-01

    This study identified the role of nuclear energy in the following three major aspects. First of all, this study carried out cost effectiveness of nuclear as a CDM technology, which is one of means of GHG emission reduction in UNFCCC. Secondly, environmental externalities caused by air pollutants emitted by power options were estimated. The 'observed market behaviour' method and 'responses to hypothetical market' method were used to estimate objectively the environmental external costs by electric source, respectively. Finally, the role of nuclear power in securing electricity supply in a liberalized electricity market was analyzed. This study made efforts to investigate whether nuclear power generation with high investment cost could be favored in a liberalized market by using 'option value' analysis of investments

  6. Nuclear integrated database and design advancement system

    International Nuclear Information System (INIS)

    Ha, Jae Joo; Jeong, Kwang Sub; Kim, Seung Hwan; Choi, Sun Young.

    1997-01-01

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs

  7. 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

  8. 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

  9. 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.; Oh, K. B.

    2002-12-01

    This study deals with current energy issues, environmental aspects of energy, project feasibility evaluation, and activities of international organizations. Current energy issues including activities related with UNFCCC, sustainable development, and global concern on energy issues were surveyed with focusing on nuclear related activities. Environmental aspects of energy includes various topics such as, inter- industrial analysis of nuclear sector, the role of nuclear power in mitigating GHG emission, carbon capture and sequestration technology, hydrogen production by using nuclear energy, Life Cycle Analysis as a method of evaluating environmental impacts of a technology, and spent fuel management in the case of introducing fast reactor and/or accelerator driven system. Project feasibility evaluation includes nuclear desalination using SMART reactor, and introduction of COMFAR computer model, developed by UNIDO to carry out feasibility analysis in terms of business attitude. Activities of international organizations includes energy planning activities of IAEA and OECD/NEA, introduction of the activities of FNCA, one of the cooperation mechanism among Asian countries. In addition, MESSAGE computer model was also introduced. The model is being developed by IAEA to effectively handle liberalization of electricity market combined with environmental constraints

  10. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR

    International Nuclear Information System (INIS)

    Jauregui Ch, V.

    2013-01-01

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X 2 but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (k eff ). The results show that the fuel lattices in which the frequency, which the inverted form of the X 2 distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of the gadolinium

  11. 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)

  12. World Nuclear Association Design Change Management

    International Nuclear Information System (INIS)

    Waddington, John Geoffrey

    2013-01-01

    This presentation treats of design change management in the regulation of nuclear fleets. It covers activities of the WNA/CORDEL/Design Change Management Task Force, including views on the roles of vendors, owner's groups, utility and design authority, WANO and the regulators. The presentation highlights differences of capabilities between large utilities with strong technical staff and smaller utilities that require support and expertise from others. It also notes the current expectation that licensees are solely responsible for the safety of the design and operation of their plants and for maintaining a full understanding and knowledge of the design within licensee's own organization in an internal entity called design authority. It encourages regulators to re-examine this expectation for design changes, arguing that while large utilities maybe be able to deal with design changes, the smaller utilities may be challenged due to their small size and lack of appropriate expertise. It further notes that the original designer must be involved in the management of design changes. In addition, the presentation emphasizes benefits of standardization in design and regulatory expectations internationally, including the benefits of increasing safety and economy. The author provides that the CORDEL Working Group uses international standardization to mean that each vendor's design can be built by a vendor, and ordered by a utility, in every country and be able to meet national regulations without significant changes other than adaptations to meet site requirements. In this discussion, he highlights the aircraft industry as an example and notes the need for internationally agreed mechanisms for design change as well as the need for formal, agreed (internationally) role for the designer to play throughout the fleet lifetime

  13. 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

  14. 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

  15. 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

  16. Nuclear instrumentation evaluation and analysis

    International Nuclear Information System (INIS)

    Park, Suk Jun; Han, Sang Joon; Chung, Chong Eun; Han, Kwang Soo; Kim, Dong Hwa; Park, Byung Hae; Moon, Je Sun; Lee, Chel Kwon; Song, Ki Sang; Choi, Myung Jin; Kim, Seung Bok; Kim, Jung Bok

    1986-12-01

    This project provides the program for improving instrumentation reliability as well as developing a cost-effective preventive maintenance activity through evaluation and analysis of nuclear instrumentation concerning pilot plants, large-scale test facilities and various laboratories on KAERI site. In addition, it discusses the program for enhancing safe operations and improving facility availability through establishment of maintenance technology. (Author)

  17. 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

  18. 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.)

  19. 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

  20. Human reliability guidance - How to increase the synergies between human reliability, human factors, and system design and engineering. Phase 2: The American Point of View - Insights of how the US nuclear industry works with human reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, J. (Vattenfall Ringhals AB, Stockholm (Sweden))

    2010-12-15

    The main goal of this Nordic Nuclear Safety Research Council (NKS) project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. The project consists of two substudies: The Nordic Point of View - A User Needs Analysis, and The American Point of View - Insights of How the US Nuclear Industry Works with HRA. The purpose of the Nordic Point of View study was a user needs analysis that aimed to survey current HRA practices in the Nordic nuclear industry, with the main focus being to connect HRA to system design. In this study, 26 Nordic (Swedish and Finnish) nuclear power plant specialists with research, practitioner, and regulatory expertise in HRA, PRA, HSI, and human performance were interviewed. This study was completed in 2009. This study concludes that HRA is an important tool when dealing with human factors in control room design or modernizations. The Nordic Point of View study showed areas where the use of HRA in the Nordic nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

  1. Human reliability guidance - How to increase the synergies between human reliability, human factors, and system design and engineering. Phase 2: The American Point of View - Insights of how the US nuclear industry works with human reliability analysis

    International Nuclear Information System (INIS)

    Oxstrand, J.

    2010-12-01

    The main goal of this Nordic Nuclear Safety Research Council (NKS) project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. The project consists of two substudies: The Nordic Point of View - A User Needs Analysis, and The American Point of View - Insights of How the US Nuclear Industry Works with HRA. The purpose of the Nordic Point of View study was a user needs analysis that aimed to survey current HRA practices in the Nordic nuclear industry, with the main focus being to connect HRA to system design. In this study, 26 Nordic (Swedish and Finnish) nuclear power plant specialists with research, practitioner, and regulatory expertise in HRA, PRA, HSI, and human performance were interviewed. This study was completed in 2009. This study concludes that HRA is an important tool when dealing with human factors in control room design or modernizations. The Nordic Point of View study showed areas where the use of HRA in the Nordic nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

  2. 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.)

  3. Visualization analysis and design

    CERN Document Server

    Munzner, Tamara

    2015-01-01

    Visualization Analysis and Design provides a systematic, comprehensive framework for thinking about visualization in terms of principles and design choices. The book features a unified approach encompassing information visualization techniques for abstract data, scientific visualization techniques for spatial data, and visual analytics techniques for interweaving data transformation and analysis with interactive visual exploration. It emphasizes the careful validation of effectiveness and the consideration of function before form. The book breaks down visualization design according to three questions: what data users need to see, why users need to carry out their tasks, and how the visual representations proposed can be constructed and manipulated. It walks readers through the use of space and color to visually encode data in a view, the trade-offs between changing a single view and using multiple linked views, and the ways to reduce the amount of data shown in each view. The book concludes with six case stu...

  4. 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)

  5. Design of a nuclear reactor cooperative controller

    International Nuclear Information System (INIS)

    Alang-Rashid, N.K.; Heger, A.S.

    1991-01-01

    This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

  6. Structural analysis of nuclear components

    International Nuclear Information System (INIS)

    Ikonen, K.; Hyppoenen, P.; Mikkola, T.; Noro, H.; Raiko, H.; Salminen, P.; Talja, H.

    1983-05-01

    THe report describes the activities accomplished in the project 'Structural Analysis Project of Nuclear Power Plant Components' during the years 1974-1982 in the Nuclear Engineering Laboratory at the Technical Research Centre of Finland. The objective of the project has been to develop Finnish expertise in structural mechanics related to nuclear engineering. The report describes the starting point of the research work, the organization of the project and the research activities on various subareas. Further the work done with computer codes is described and also the problems which the developed expertise has been applied to. Finally, the diploma works, publications and work reports, which are mainly in Finnish, are listed to give a view of the content of the project. (author)

  7. 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

  8. 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

  9. Thermal and flow analysis of the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% effort of Title 1)

    International Nuclear Information System (INIS)

    Steinke, R.G.; Mueller, C.; Knight, T.D.

    1998-03-01

    The computational fluid dynamics code CFX4.2 was used to evaluate steady-state thermal-hydraulic conditions in the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% of Title 1). Thirteen facility cases were evaluated with varying temperature dependence, drywell-array heat-source magnitude and distribution, location of the inlet tower, and no-flow curtains in the drywell-array vault. Four cases of a detailed model of the inlet-tower top fixture were evaluated to show the effect of the canopy-cruciform fixture design on the air pressure and flow distributions

  10. Design of nuclear instruments for industrial use

    International Nuclear Information System (INIS)

    Maggio, G.E.

    1988-01-01

    Following an introduction to the atomic structure and the radioactive desintegration, the applications of radioisotopic sealed sources are described. The laws that govern the interaction of radiation with matter and the statistics applied to the radioactive measurements are presented. Different measurement techniques, basic equations of design, the way to provide the activity calculation of a source and the detector's characteristics are given, according to the parameters to be measured and the conditions imposed. Finally, the principles of operation and the most important characteristics of different nuclear instruments to be used in industrial measurements are described. (Author) [es

  11. 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.

  12. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  13. 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

  14. 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.

  15. 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

  16. Design and construction of nuclear power plants

    International Nuclear Information System (INIS)

    Meiswinkel, Ruediger; Meyer, Julian; Schnell, Juergen

    2013-01-01

    Despite all the efforts being put into expanding renewable energy sources, large-scale power stations will be essential as part of a reliable energy supply strategy for a longer period. Given that they are low on CO2 emissions, many countries are moving into or expanding nuclear energy to cover their baseload supply. Building structures required for nuclear installations whose protective function means they are classified as safety-related, have to meet particular construction requirements more stringent than those involved in conventional construction. This book gives a comprehensive overview from approval aspects given by nuclear and construction law, with special attention to the interface between plant and construction engineering, to a building structure classification. All life cycle phases are considered, with the primary focus on execution. Accidental actions on structures, the safety concept and design and fastening systems are exposed to a particular treatment. Selected chapters of the German concrete yearbook ''Beton-Kalender'' are now available in English. The new English BetonKalender Series delivers internationally useful engineering expertise and industrial know-how from Germany.

  17. 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)

  18. 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

  19. Programs for nuclear data analysis

    International Nuclear Information System (INIS)

    Bell, R.A.I.

    1975-01-01

    The following report details a number of programs and subroutines which are useful for analysis of data from nuclear physics experiments. Most of them are available from pool pack 005 on the IBM1800 computer. All of these programs are stored there as core loads, and the subroutines and functions in relocatable format. The nature and location of other programs are specified as appropriate. (author)

  20. Prompt nuclear analysis bibliography 1976

    International Nuclear Information System (INIS)

    Bird, J.R.; Campbell, B.L.; Cawley, R.J.

    1978-05-01

    A prompt nuclear analysis bibliography published in 1974 has been updated to include literature up to the end of 1976. The number of publications has more than doubled since mid-1973. The bibliography is now operated as a computer file and searches can be made on key words and parameters. Tables of references are given for each of the categories: backscattering, ion-ion, ion-gamma, ion-neutron, neutron-gamma, neutron-neutron and gamma-ray-induced reactions

  1. Aseismic design of Hamaoka Nuclear Power Plant

    International Nuclear Information System (INIS)

    Mizuno, Norihiro

    1975-01-01

    The Hamaoka Nuclear Power Plant of Chubu Electric Power Co. is designed so as to maintain structural safety against an earthquake of 300 gal. For the purpose, a compound-type reactor-housing building is employed, which contains a reactor, operation control and waste disposal facilities. The merits accruing from this scheme are as follows. (1) The shielding walls of the waste disposal facility can be utilized effectively in aseismatic design, leading to the increased rigidity of the building and the uniform distribution of resistance. (2) Due to the large area of the foundation, the load in earthquake can be mitigated, and it resulted in the higher structural stability. Moreover, seismic energy can be dissipated into ground. After the description of the compound building structure, it is explained how the structural resistance and the ground dissipation of seismic energy contribute to potential earthquake resistance. (Mori, K.)

  2. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  3. 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

  4. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Turgut, M. H.

    2009-01-01

    Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

  5. 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)

  6. Economic analysis of nuclear energy

    International Nuclear Information System (INIS)

    Lee, Man Ki; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Song, K. D.; Oh, K. B.

    2004-12-01

    This study evaluated the role of nuclear energy in various aspects in order to provide a more comprehensive standard of judgement to the justification of the utilization of nuclear energy. Firstly, this study evaluated the economic value addition of nuclear power generation technology and Radio-Isotope(RI) technology quantitatively by using modified Input-Output table. Secondly, a comprehensive cost-benefit analysis of nuclear power generation was conducted with an effort to quantify the foreign exchange expenditure, the environmental damage cost during 1986-2015 for each scenario. Thirdly, the effect of the regulation of CO 2 emission on the Korean electric supply system was investigated. In more detail, an optimal composition of power plant mix by energy source was investigated, under the assumption of the CO 2 emission regulation at a certain level, by using MESSAGE model. Finally, the economic spillover effect from technology self-reliance of NSSS by Korea Atomic Energy Research Institute was evaluated. Both production spillover effect and value addition spillover effect were estimated by using Input-Output table

  7. 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

  8. Project designing of Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Krychtalek, Z.; Linek, V.

    1989-01-01

    The geological and seismic parameters are listed of the Temelin nuclear power plant. The division of the site in building zones is described. The main zones consist of the power generation unit zone with the related auxiliary buildings of hot plants and of the auxiliary buildings of the nonactive part with industrial buildings. The important buildings are interconnected with communication and technology bridges. Cooling towers and spray pools and the entrance area are part of the urbanistic design. The architectonic design of the buildings uses standard building elements and materials. The design of the buildings is based on the requirements on their function and on structural load and on the demands of maximal utilization of the type of the reinforced concrete prefab structure system. The structure is made of concrete or steel cells. The project design is based on Soviet projects. The layout is shown of the main power generation units and a section is presented of a 1,000 MW unit. (J.B.). 2 figs

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. Nuclear island buildings layout collaborative design platform development and application

    International Nuclear Information System (INIS)

    Chen Li; Huang Wenqiang

    2014-01-01

    Based on characteristics of nuclear island layout design, the large number files and complicated interface, for realizing collaborative design and fine management goal, Establish collaborative design platform, which includes the design task module, 3D design module, project management module. These three modules can package design input files, realize synchronous design and real-time track design drawings state, timely feedback between design, procurement, construction site. There is no design task delay due to tracking and has realized fine management of design. (authors)

  15. 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)

  16. 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

  17. Lunar mission design using nuclear thermal rockets

    International Nuclear Information System (INIS)

    Stancati, M.L.; Collins, J.T.; Borowski, S.K.

    1991-01-01

    The NERVA-class Nuclear Thermal Rocket (NTR), with performance nearly double that of advanced chemical engines, has long been considered an enabling technology for human missions to Mars. NTR engines address the demanding trip time and payload delivery needs of both cargo-only and piloted flights. But NTR can also reduce the Earth launch requirements for manned lunar missions. First use of NTR for the Moon would be less demanding and would provide a test-bed for early operations experience with this powerful technology. Study of application and design options indicates that NTR propulsion can be integrated with the Space Exploration Initiative scenarios to deliver performance gains while managing controlled, long-term disposal of spent reactors to highly stable orbits

  18. Design of a nuclear steam reforming plant

    International Nuclear Information System (INIS)

    Malherbe, J.

    1980-01-01

    The design of a plant for the steam reforming of methane using a High Temperature Reactor has been studied by CEA in connection with the G.E.G.N. This group of companies (CEA, GAZ DE FRANCE, CHARBONNAGES DE FRANCE, CREUSOT-LOIRE, NOVATOME) is in charge of studying the feasibility of the coal gasification process by using a nuclear reactor. The process is based on the hydrogenation of the coal in liquid phase with hydrogen produced by a methane steam reformer. The reformer plant is fed by a pipe of natural gas or SNG. The produced hydrogen feeds the gasification plant which could not be located on the same site. An intermediate hydrogen storage between the two plants could make the coupling more flexible. The gasification plant does not need a great deal of heat and this heat can be satisfied mostly by internal heat exchanges

  19. 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

  20. 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

  1. 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)

  2. Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

    International Nuclear Information System (INIS)

    Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

    2004-01-01

    Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

  3. Nuclear Proliferation Technology Trends Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, Michael D.; Coles, Garill A.; Talbert, Robert J.

    2005-10-04

    A process is underway to develop mature, integrated methodologies to address nonproliferation issues. A variety of methodologies (both qualitative and quantitative) are being considered. All have one thing in common, a need for a consistent set of proliferation related data that can be used as a basis for application. One approach to providing a basis for predicting and evaluating future proliferation events is to understand past proliferation events, that is, the different paths that have actually been taken to acquire or attempt to acquire special nuclear material. In order to provide this information, this report describing previous material acquisition activities (obtained from open source material) has been prepared. This report describes how, based on an evaluation of historical trends in nuclear technology development, conclusions can be reached concerning: (1) The length of time it takes to acquire a technology; (2) The length of time it takes for production of special nuclear material to begin; and (3) The type of approaches taken for acquiring the technology. In addition to examining time constants, the report is intended to provide information that could be used to support the use of the different non-proliferation analysis methodologies. Accordingly, each section includes: (1) Technology description; (2) Technology origin; (3) Basic theory; (4) Important components/materials; (5) Technology development; (6) Technological difficulties involved in use; (7) Changes/improvements in technology; (8) Countries that have used/attempted to use the technology; (9) Technology Information; (10) Acquisition approaches; (11) Time constants for technology development; and (12) Required Concurrent Technologies.

  4. 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

  5. Risk and safety analysis of nuclear systems

    National Research Council Canada - National Science Library

    Lee, John C; McCormick, Norman J

    2011-01-01

    ...), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear applications, although there is an emphasis placed on the analysis of nuclear systems...

  6. 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)

  7. 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

  8. Value of life and nuclear design

    International Nuclear Information System (INIS)

    Howard, R.A.; Matheson, J.E.; Owen, D.L.

    1978-01-01

    Using the decision analysis framework, an implicit value of life can be determined for design decisions that have death as a possible outcome. A survey of the literature and calculations suggest a great inconsistency in the implied value of life. It is shown that by using a consistent, explicit value of life, the total expected number of deaths from all projects can be reduced without increasing total expenditures or reducing benefits. The explicit value of life directly affects design decisions. Some recent research indicates that the value that an individual places on his own life can be characterized by a few assessments of the individual's circumstances and his preferences

  9. DESIGN INFORMATION VERIFICATION FOR NUCLEAR SAFEGUARDS

    Energy Technology Data Exchange (ETDEWEB)

    Robert S. Bean; Richard R. M. Metcalf; Phillip C. Durst

    2009-07-01

    A critical aspect of international safeguards activities performed by the International Atomic Energy Agency (IAEA) is the verification that facility design and construction (including upgrades and modifications) do not create opportunities for nuclear proliferation. These Design Information Verification activities require that IAEA inspectors compare current and past information about the facility to verify the operator’s declaration of proper use. The actual practice of DIV presents challenges to the inspectors due to the large amount of data generated, concerns about sensitive or proprietary data, the overall complexity of the facility, and the effort required to extract just the safeguards relevant information. Planned and anticipated facilities will (especially in the case of reprocessing plants) be ever larger and increasingly complex, thus exacerbating the challenges. This paper reports the results of a workshop held at the Idaho National Laboratory in March 2009, which considered technologies and methods to address these challenges. The use of 3D Laser Range Finding, Outdoor Visualization System, Gamma-LIDAR, and virtual facility modeling, as well as methods to handle the facility data issues (quantity, sensitivity, and accessibility and portability for the inspector) were presented. The workshop attendees drew conclusions about the use of these techniques with respect to successfully employing them in an operating environment, using a Fuel Conditioning Facility walk-through as a baseline for discussion.

  10. 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

  11. 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)

  12. 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

  13. 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.

  14. Nuclear Power Plant Module, NPP-1: Nuclear Power Cost Analysis.

    Science.gov (United States)

    Whitelaw, Robert L.

    The purpose of the Nuclear Power Plant Modules, NPP-1, is to determine the total cost of electricity from a nuclear power plant in terms of all the components contributing to cost. The plan of analysis is in five parts: (1) general formulation of the cost equation; (2) capital cost and fixed charges thereon; (3) operational cost for labor,…

  15. Parity simulation for nuclear plant analysis

    International Nuclear Information System (INIS)

    Hansen, K.F.; Depiente, E.

    1986-01-01

    The analysis of the transient performance of nuclear plants is sufficiently complex that simulation tools are needed for design and safety studies. The simulation tools are needed for design and safety studies. The simulation tools are normally digital because of the speed, flexibility, generality, and repeatability of digital computers. However, communication with digital computers is an awkward matter, requiring special skill or training. The designer wishing to gain insight into system behavior must expend considerable effort in learning to use computer codes, or else have an intermediary communicate with the machine. There has been a recent development in analog simulation that simplifies the user interface with the simulator, while at the same time improving the performance of analog computers. This development is termed parity simulation and is now in routine use in analyzing power electronic network transients. The authors describe the concept of parity simulation and present some results of using the approach to simulate neutron kinetics problems

  16. 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.

  17. 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.

  18. Designing a mini subcritical nuclear reactor

    International Nuclear Information System (INIS)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

    2015-10-01

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of 239 PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of 239 PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k e -f f , the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k eff , the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons 239 PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k eff was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k eff of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five cylinders not met. (Author)

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. Analysis of Design Documentation

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp

    1998-01-01

    In design practice a process where a satisfactory solution is created within limited resources is required. However, since the design process is not well understood, research into how engineering designers actually solve design problems is needed. As a contribution to that end a research project...

  4. 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)

  5. Nuclear plant analyzer development and analysis applications

    International Nuclear Information System (INIS)

    Laats, E.T.

    1984-01-01

    The Nuclear Plant Analyzer (NPA) is being developed as the U.S. Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. The NPA integrates the NRC's computerized reactor behavior simulation codes such as RELAP5 and TRAC-BWR, both of which are well-developed computer graphics programs, and large repositories of reactor design and experimental data. Utilizing the complex reactor behavior codes as well as the experiment data repositories enables simulation applications of the NPA that are generally not possible with more simplistic, less mechanistic reactor behavior codes. These latter codes are used in training simulators or with other NPA-type software packages and are limited to displaying calculated data only. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

  6. Complete analysis of a nuclear building to nuclear safety standards

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, T A

    1975-01-01

    The nuclear standards impose on the designer the necessity of examining the loads, stresses, and strains in a nuclear building even under extreme loading conditions, both due to plant malfunctions and environmental accidents. It is necessary then to generate, combine, and examine a tremendous amount of data; really the lack of symmetry and general complication of the structures and the large number of loading combinations make an automatic analysis quite necessary. A largely automated procedure is presented in view of solving the problem by a series of computer programs linked together. After the seismic analysis has been performed by (SADE CODE) these data together with the data coming from thermal specifications, weight, accident descriptions etc. are fed into a finite element computer code (SAP4) for analysis. They are processed and combined by a computer code (COMBIN) according to the loading conditions (the usual list in Italy is given and briefly discussed), so that for each point (or each selected zone) under each loading condition the applied loads are listed. These data are fed to another computer code (DTP), which determines the amount of reinforcing bars necessary to accommodate the most severe of the loading conditions. The Aci 318/71 and Italian regulation procedures are followed; the characteristics of the program are briefly described and discussed. Some particular problems are discussed, e.g. the thermal stresses due to normal and accident conditions, the inelastic behavior of some frame elements (due to concrete cracking) is considered by means of an 'ad hoc' code. Typical examples are presented and the results are discussed showing a relatively large benefit in considering this inelastic effect.

  7. Requirements for nuclear data from ITER/FER nuclear design

    International Nuclear Information System (INIS)

    Maki, Koichi

    1991-01-01

    Considering ITER and FER activities and future programme of fusion reactor developments, the present situations in fusion neutronics were explained up to now from the previous fusion nuclear data specialist meeting. Vicissitude of development of FUSION-J3 was also explained. From these discussions we clarified the required accuracies in nuclear data, their time limits and their priorities of nuclides to be improved for fusion reactor developments. (author)

  8. 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)

  9. 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)

  10. Integral design small nuclear power plant UNITHERM

    International Nuclear Information System (INIS)

    Adamovich, L. A.; Grechko, G. I.; Ulasevich, V. K.; Shishkin, V. A.

    1995-01-01

    The need to erect expensive energy transmission lines to these places demands to use independent local energy sources. Therefore, a reasonable alternative to the plants fired fossil fuel, mostly hydrocarbon fuel, may come from the nuclear power plants (NPP) of relatively small capacity which are nonattended, shipped to the site by large-assembled modules and completely withdrawable from the site during decommissioning. Application of NPPs for power and heat supply may prove to be cost-efficient and rather positive from social and ecological point of view. UNITHERM NPP belongs to such energy sources and may be used for heat and power supply. Heat can be provided both as hot water and superheated steam. The consumers are able to specify heat/energy supply ratio. NPP design provides for independent energy supply to the consumers and the possibility to disconnect each of them without disruption of operation of the others. Thermal hydraulic diagram of UNITHERM NPP provides for the use of three interconnected, process circuits. The consumers of thermal energy (turbogenerator unit and boilers of the central heating unit) are arranged in the last circuit

  11. 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

  12. Analysis of india and Pakistan's nuclear capacity

    International Nuclear Information System (INIS)

    Li Zhimin

    1999-07-01

    The development and capacity of both India and Pakistan's nuclear weapons are described in production of weapon-grade materials, nuclear testing, weaponization engineering and delivery systems. India is capable of designing and manufacturing both small yield tactic nuclear weapons and big yield strategic ones and also possesses the technique to design and manufacture H-bombs. Weapon-grade plutonium constitutes the primary fission material for India's nuclear weapon and it has plutonium enough to make 70 to 100 nuclear weapons. India can also produce some tritium. India has already possessed delivery systems but it has not yet mounted nuclear warheads on its ballistic missiles even though its missiles, which India has already owned or is under development, have the ability to carry nuclear warheads. Pakistan also has the ability to make both tactic nuclear weapons and strategic ones. With its weapon-grade uranium, 20 to 30 nuclear weapons can be made. Besides the uranium production facility. Pakistan also has the facility to produce tritium. It is supposed that Pakistan has the ability to carry nuclear weapons with airplane, but it has a long way to go if it wants to mount nuclear weapon, especially bit yield ones, on its own missile. As a whole, India's nuclear force is stronger than Pakistan's, and its development far more advanced than Pakistan's

  13. 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

  14. Nondestructive assay methodologies in nuclear forensics analysis

    International Nuclear Information System (INIS)

    Tomar, B.S.

    2016-01-01

    In the present chapter, the nondestructive assay (NDA) methodologies used for analysis of nuclear materials as a part of nuclear forensic investigation have been described. These NDA methodologies are based on (i) measurement of passive gamma and neutrons emitted by the radioisotopes present in the nuclear materials, (ii) measurement of gamma rays and neutrons emitted after the active interrogation of the nuclear materials with a source of X-rays, gamma rays or neutrons

  15. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  16. 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

  17. 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.

  18. Applications of noise analysis to nuclear safety

    International Nuclear Information System (INIS)

    Aguilar Martinez, Omar

    2000-01-01

    Noise Analysis techniques (analysis of the fluctuation of physical parameters) have been successfully applied to the operational vigilance of the technical equipment that plays a decisive role in the production cycle of a very complex industry. Although fluctuation measurements in nuclear installations started almost at the start of the nuclear era (see works by Feynman and Rossi on the development of neutron methodology), only recently have neutron noise diagnostic applications begun to be a part of the standard procedures for the performance of some modern nuclear installations. Following the relevant technical advances made in information sciences and analogical electronics, measuring the fluctuation of physical parameters has become a very effective tool for detecting, guarding and following up possible defects in a nuclear system. As the processing techniques for the fluctuation of a nuclear reactor's physical-neutron parameters have evolved (temporal and frequency analysis, multi-parameter self -regression analysis, etc.), the applications of the theory of non-lineal dynamics and chaos theory have progressed by focusing on the problem from another perspective. This work reports on those nuclear applications of noise analysis that increase nuclear safety in all types of nuclear facilities and that have been carried out by the author over the last decade, such as: -Void Force Critical Set Applications (Zero Power Reactor Applications, Central Institute of Physical Research, Budapest, Hungary); -Research Reactor Applications (Triga Mark III Reactor, National Institute of Nuclear Research, ININ, Mexico); -Power Reactor Applications in a Nuclear Power Plant (First Circuit of Block II, Paks Nuclear Center, Hungary); -Second Loop applications in a Nuclear Power Plant (Block I Paks Nuclear Center, Hungary; Block II Kalinin Nuclear Center, Russia); -Shield System Applications for the Transport of Radioisotopes (Nuclear Technology Center, Havana, Cuba) New trends in

  19. Nuclear spectrometry for environmental analysis and mapping

    International Nuclear Information System (INIS)

    Simon, Aliz

    2013-01-01

    Full text: The International Atomic Energy Agency (IAEA) helps countries to mobilize peaceful applications of nuclear science and technology. The three main pillars of the activities are: safety and security; science and technology; and safeguards and verification. As part of the science and technology pillar, the Physics Section supports Member States regarding utilization of particle accelerators and research reactors, applications of nuclear instrumentation, and controlled nuclear fusion research. Support is provided to the Member States in the form of capacity building, knowledge transfer and networking. The IAEA's coordinated research activities are designed to contribute to this mandate, by stimulating and coordinating research in IAEA Member States in selected nuclear fields. These coordinated research activities are normally implemented through Coordinated Research Projects that bring together research institutes from both developing and developed Member States to collaborate on the research topic of interest. The establishment of sustainable education and training programmes is fundamental for the safe, secure and efficient development of the nuclear field. The lAEA offers a wide spectrum of activities in support of education, training, human resource development and capacity building including interregional, regional and national training courses and workshops; assists visits and reviews services; initiates, formulates and runs programmes; networks managers and specialists for sharing good practices; assists in publications that compile the best international practices; supplies training materials and training tools; and supports internship programmes for the young generations of scientists and fellows. For the developing countries, the Technical Cooperation Programme provides the necessary skills and equipment to establish sustainable technology in the counterpart country or region through training courses, expert missions, fellowships, scientific

  20. Nuclear spectrometry for environmental analysis and mapping

    Energy Technology Data Exchange (ETDEWEB)

    Simon, Aliz, E-mail: Aliz.Simon@iaea.org [International Atomic Energy Agency (IAEA), Division of Physical and Chemical Sciences, Vienna (Austria)

    2013-07-01

    Full text: The International Atomic Energy Agency (IAEA) helps countries to mobilize peaceful applications of nuclear science and technology. The three main pillars of the activities are: safety and security; science and technology; and safeguards and verification. As part of the science and technology pillar, the Physics Section supports Member States regarding utilization of particle accelerators and research reactors, applications of nuclear instrumentation, and controlled nuclear fusion research. Support is provided to the Member States in the form of capacity building, knowledge transfer and networking. The IAEA's coordinated research activities are designed to contribute to this mandate, by stimulating and coordinating research in IAEA Member States in selected nuclear fields. These coordinated research activities are normally implemented through Coordinated Research Projects that bring together research institutes from both developing and developed Member States to collaborate on the research topic of interest. The establishment of sustainable education and training programmes is fundamental for the safe, secure and efficient development of the nuclear field. The lAEA offers a wide spectrum of activities in support of education, training, human resource development and capacity building including interregional, regional and national training courses and workshops; assists visits and reviews services; initiates, formulates and runs programmes; networks managers and specialists for sharing good practices; assists in publications that compile the best international practices; supplies training materials and training tools; and supports internship programmes for the young generations of scientists and fellows. For the developing countries, the Technical Cooperation Programme provides the necessary skills and equipment to establish sustainable technology in the counterpart country or region through training courses, expert missions, fellowships, scientific

  1. Application of nuclear activation analysis

    International Nuclear Information System (INIS)

    Mamonov, E.I.; Khlystova, A.F.

    1979-01-01

    Consideration is given to the applications of nuclear-activation analysis (NAA) as discussed at the International Conference of 1977. One of the new results in the present-day NAA practices is the growing number of elements detected in samples without using a destructive radiochemical separation. An essential feature in this context is the development of the system automation of control and information NAA operations through the use computers. In biological medicine a multicomponent NAA is employed to determine the concentration of elements in various human organs and objects, in metabolic studies and for diagnostic purposes. In agriculture NAA finds applications in the evaluation of grain protein, analysis of element feed composition, soil and fertilizers. The application of this method to the environmental monitoring is considered with particular reference to the element analysis of water (especially drinking water), air, plant residues. Data are presented for the use of NAA in metallurgy, geology, archaeology and criminal law. Tables are provided to illustrate the uses of NAA in various fields

  2. 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)

  3. 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.)

  4. 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)

  5. 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)

  6. 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

  7. 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.

  8. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  9. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    2016-08-26

    Aug 26, 2016 ... The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  10. Integrating reliability analysis and design

    International Nuclear Information System (INIS)

    Rasmuson, D.M.

    1980-10-01

    This report describes the Interactive Reliability Analysis Project and demonstrates the advantages of using computer-aided design systems (CADS) in reliability analysis. Common cause failure problems require presentations of systems, analysis of fault trees, and evaluation of solutions to these. Results have to be communicated between the reliability analyst and the system designer. Using a computer-aided design system saves time and money in the analysis of design. Computer-aided design systems lend themselves to cable routing, valve and switch lists, pipe routing, and other component studies. At EG and G Idaho, Inc., the Applicon CADS is being applied to the study of water reactor safety systems

  11. 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)

  12. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors; Calificacion del programa WIMS de calculo neutronico para diseno, seguimiento de operacion y analisis de accidentes de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M [Ente Nacional Regulador Nuclear, Buenos Aires (Argentina)

    1997-12-31

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

  13. Palo Verde nuclear dynamic analysis (PANDA)

    International Nuclear Information System (INIS)

    Girjashankar, P.V.; Secker, P.A. Jr.; LeClair, S.J.; Mendoza, J.; Webb, J.R.

    1988-01-01

    Arizona Nuclear Power Project (ANPP) has initiated the development of a large scale dynamic analysis computer program for the Palo Verde Nuclear Generating Station (PVNGS). This paper presents the decision processes and preliminary development activities that have been pursued related to the code development. The PANDA (Palo Verde Nuclear Dynamic Analysis) code will be used for a variety of applications as described in this paper

  14. 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

  15. 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.

  16. Risk analysis of nuclear safeguards regulations

    International Nuclear Information System (INIS)

    Al-Ayat, R.A.; Altman, W.D.; Judd, B.R.

    1982-06-01

    The Aggregated Systems Model (ASM), a probabilisitic risk analysis tool for nuclear safeguards, was applied to determine benefits and costs of proposed amendments to NRC regulations governing nuclear material control and accounting systems. The objective of the amendments was to improve the ability to detect insiders attempting to steal large quantities of special nuclear material (SNM). Insider threats range from likely events with minor consequences to unlikely events with catastrophic consequences. Moreover, establishing safeguards regulations is complicated by uncertainties in threats, safeguards performance, and consequences, and by the subjective judgments and difficult trade-offs between risks and safeguards costs. The ASM systematically incorporates these factors in a comprehensive, analytical framework. The ASM was used to evaluate the effectiveness of current safeguards and to quantify the risk of SNM theft. Various modifications designed to meet the objectives of the proposed amendments to reduce that risk were analyzed. Safeguards effectiveness was judged in terms of the probability of detecting and preventing theft, the expected time to detection, and the expected quantity of SNM diverted in a year. Data were gathered in tours and interviews at NRC-licensed facilities. The assessment at each facility was begun by carefully selecting scenarios representing the range of potential insider threats. A team of analysts and facility managers assigned probabilities for detection and prevention events in each scenario. Using the ASM we computed the measures of system effectiveness and identified cost-effective safeguards modifications that met the objectives of the proposed amendments

  17. 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

  18. 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)

  19. 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

  20. Design and realisation of a microprogramme for the analysis of nuclear spectroscopic data on the intelligent terminal, H.P.2648A

    International Nuclear Information System (INIS)

    Tendeku, F.K.

    1980-01-01

    A microprogramme has been developed for the analysis of nuclear spectroscopic data on the microprocessor-controlled terminal, H.P. 2648A. The terminal enables data to be processed locally. Many operations normally requiring connection to a computer can be performed on a stand alone basis. The principal aspects of the microprogramme are the graphical display of spectral data, the automatic and manual extraction of peaks, and the determination of peak characteristics. Among the features of the microprogramme are algorithms which enable a rapid graphical display of data in logarithmic and linear scales and in divers modes as well as functions enabling image amplification. The automatic peak searching algorithm represents a modified form of the Mariscotti method. The peak analysis procedure offers two options. The first makes use of the data directly. The second is based on the determination of the parameters of the gaussian function which best fits the data points, using the non-linear least squares principle. The terminal has been successfully interfaced to an IBM 370 computer through a satellite computer using the Parallel Duplex Register module of Hewlett Packard [fr

  1. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  2. Request from nuclear fuel cycle and criticality safety design

    International Nuclear Information System (INIS)

    Hamasaki, Manabu; Sakashita, Kiichiro; Natsume, Toshihiro

    2005-01-01

    The quality and reliability of criticality safety design of nuclear fuel cycle systems such as fuel fabrication facilities, fuel reprocessing facilities, storage systems of various forms of nuclear materials or transportation casks have been largely dependent on the quality of criticality safety analyses using qualified criticality calculation code systems and reliable nuclear data sets. In this report, we summarize the characteristics of the nuclear fuel cycle systems and the perspective of the requirements for the nuclear data, with brief comments on the recent issue about spent fuel disposal. (author)

  3. 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

  4. 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.

  5. 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.

  6. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-12-01

    In nuclear or shielding design analysis for reactors including nuclear facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multigroup constant library using the newly compiled data files and the code systems. As the results of this project, JEF-2.2 which is latest version of Joint Evaluated File developed at OECD/NEA was compiled and COMPLOT and EVALPLOT utility codes were installed in personal computer, which are able to draw ENDF/B-formatted nuclear data for comparison and check. Computer system (NJOY/ACER) for generating continuous energy Monte Carlo code MCNP library was established and the system was validated by analyzing a number of experimental data. (Author).

  7. Nuclear analysis methods. Rudiments of radiation protection

    International Nuclear Information System (INIS)

    Roth, E.

    1998-01-01

    The nuclear analysis methods are generally used to analyse radioactive elements but they can be used also for chemical analysis, with fields such analysis and characterization of traces. The principles of radiation protection are explained (ALARA), the biological effects of ionizing radiations are given, elements and units used in radiation protection are reminded in tables. A part of this article is devoted to how to use radiation protection in a nuclear analysis laboratory. (N.C.)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. Economic analysis of nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S

    1999-12-01

    The objective of this study is to analyze how the economics of nuclear power generation are affected by the change in nuclear environmental factors and then, to suggest desirable policy directions to improve the efficiency of the use of nuclear energy resources in korea. This study focused to analyze the impact of the change in 3 major nuclear environmental factors in Korea on the economics of nuclear power generation. To do this, environmental external cost, nuclear R and fund, and carbon emission control according to UNFCCC were selected as the major factors. First of all, this study evaluated the impacts on the health and the environment of air pollutants emitted from coal power plant and nuclear power plant, two major electric power generating options in Korea. Then, the environmental external costs of those two options were estimated by transforming the health and environmental impact in to monetary values. To do this, AIRPACTS and 'Impacts of atmospheric release' model developed by IAEA were used. Secondly, the impact of nuclear R and D fund raised by the utility on the increment of nuclear power generating cost was evaluated. Then, the desirable size of the fund in Korea was suggested by taking into consideration the case of Japan. This study also analyzed the influences of the fund on the economics of nuclear power generation. Finally, the role of nuclear power under the carbon emission regulation was analyzed. To do this, the econometric model was developed and the impact of the regulation on the national economy was estimated. Further efforts were made to estimate the role by developing CGE model in order to improve the reliability of the results from the econometric model.

  16. Economic analysis of nuclear energy

    International Nuclear Information System (INIS)

    Song, Ki Dong; Lee, M. K.; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Kim, H. S.

    1999-12-01

    The objective of this study is to analyze how the economics of nuclear power generation are affected by the change in nuclear environmental factors and then, to suggest desirable policy directions to improve the efficiency of the use of nuclear energy resources in korea. This study focused to analyze the impact of the change in 3 major nuclear environmental factors in Korea on the economics of nuclear power generation. To do this, environmental external cost, nuclear R and fund, and carbon emission control according to UNFCCC were selected as the major factors. First of all, this study evaluated the impacts on the health and the environment of air pollutants emitted from coal power plant and nuclear power plant, two major electric power generating options in Korea. Then, the environmental external costs of those two options were estimated by transforming the health and environmental impact in to monetary values. To do this, AIRPACTS and 'Impacts of atmospheric release' model developed by IAEA were used. Secondly, the impact of nuclear R and D fund raised by the utility on the increment of nuclear power generating cost was evaluated. Then, the desirable size of the fund in Korea was suggested by taking into consideration the case of Japan. This study also analyzed the influences of the fund on the economics of nuclear power generation. Finally, the role of nuclear power under the carbon emission regulation was analyzed. To do this, the econometric model was developed and the impact of the regulation on the national economy was estimated. Further efforts were made to estimate the role by developing CGE model in order to improve the reliability of the results from the econometric model

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of 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 the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  3. 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

  4. Economic analysis of nuclear power generation

    International Nuclear Information System (INIS)

    Song, Ki Dong; Choi, Young Myung; Kim, Hwa Sup; Lee, Man Ki; Moon, Kee Hwan; Kim, Seung Su

    1997-12-01

    The major contents in this study are as follows : - long-term forecast to the year of 2040 is provided for nuclear electricity generating capacity by means of logistic curve fitting method. - the role of nuclear power in a national economy is analyzed in terms of environmental regulation. To do so, energy-economy linked model is developed. By using this model, the benefits from the introduction of nuclear power in Korea are estimated. Study on inter-industry economic activity for nuclear industry is carried out by means of an input-output analysis. Nuclear industry is examined in terms of inducement effect of production, of value-added, and of import. - economic analysis of nuclear power generation is performed especially taking into consideration wide variations of foreign currency exchange rate. The result is expressed in levelized generating costs. (author). 27 refs., 24 tabs., 44 figs

  5. Simplified design of flexible expansion anchored plates for nuclear structures

    International Nuclear Information System (INIS)

    Mehta, N.K.; Hingorani, N.V.; Longlais, T.G.; Sargent and Lundy, Chicago, IL)

    1984-01-01

    In nuclear power plant construction, expansion anchored plates are used to support pipe, cable tray and HVAC duct hangers, and various structural elements. The expansion anchored plates provide flexibility in the installation of field-routed lines where cast-in-place embedments are not available. General design requirements for expansion anchored plate assemblies are given in ACI 349, Appendix B (1). The manufacturers recommend installation procedures for their products. Recent field testing in response to NRC Bulletin 79-02 (2) indicates that anchors, installed in accordance with manufacturer's recommended procedures, perform satisfactorily under static and dynamic loading conditions. Finite element analysis is a useful tool to correctly analyze the expansion anchored plates subject to axial tension and biaxial moments, but it becomes expensive and time-consuming to apply this tool for a large number of plates. It is, therefore, advantageous to use a simplified method, even though it may be more conservative as compared to the exact method of analysis. This paper presents a design method referred to as the modified rigid plate analysis approach to simplify both the initial design and the review of as-built conditions

  6. Design principles of a nuclear and industrial HVAC of IFMIF

    International Nuclear Information System (INIS)

    Pruneri, Giuseppe; Ibarra, A.; Heidinger, R.; Knaster, J.; Sugimoto, M.

    2016-01-01

    Highlights: • Parameter of Derivate air Contamination (DAC) allows to associate the type of air ventilation. • The construction and operation of IFMIF will be subjected to the regulations of the country in which it will be sited. • Structures, systems and components are assigned a particular safety important components (SIC, 1–2 and Non-SIC) clarification that is based on the consequences of their failure. • Reliability, Availability, Maintainability and Inspectability (RAMI) analysis has given a great contribution of the facility to optimize the configuration, particularly for the HVAC system. - Abstract: In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1]. The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2]. The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their

  7. Design principles of a nuclear and industrial HVAC of IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Pruneri, Giuseppe [IFMIF/EVEDA, Project Team, Rokkasho (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Heidinger, R. [F4E, Garching (Germany); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho (Japan); Sugimoto, M. [JAEA, Rokkasho (Japan)

    2016-02-15

    Highlights: • Parameter of Derivate air Contamination (DAC) allows to associate the type of air ventilation. • The construction and operation of IFMIF will be subjected to the regulations of the country in which it will be sited. • Structures, systems and components are assigned a particular safety important components (SIC, 1–2 and Non-SIC) clarification that is based on the consequences of their failure. • Reliability, Availability, Maintainability and Inspectability (RAMI) analysis has given a great contribution of the facility to optimize the configuration, particularly for the HVAC system. - Abstract: In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1]. The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2]. The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their

  8. Computational Design of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-01-01

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  9. Design and construction of nuclear power plants

    CERN Document Server

    Schnell, Jürgen; Meiswinkel, Rüdiger; Bergmeister, Konrad; Fingerloos, Frank; Wörner, Johann-Dietrich

    2013-01-01

    Despite all the efforts being put into expanding renewable energy sources, large-scale power stations will be essential as part of a reliable energy supply strategy for a longer period. Given that they are low on CO2 emissions, many countries are moving into or expanding nuclear energy to cover their baseload supply.Building structures required for nuclear installations whose protective function means they are classified as safety-related, have to meet particular construction requirements more stringent than those involved in conventional construction. This book gives a comprehensive overv

  10. Design of blast simulators for nuclear testing

    International Nuclear Information System (INIS)

    Mark, A.; Opalka, K.O.; Kitchens, C.W. Jr.

    1983-01-01

    A quasi-one-dimensional computational technique is used to model the flow of a large, complicated shock tube. The shock tube, or Large Blast Simulator, is used to simulate conventional or nuclear explosions by shaping the pressure history. Results from computations show favorable agreement when compared with data taken in the facility at Gramat, France. Such future shock tubes will include a thermal irradiation capability to better simulate a nuclear event. The computations point to the need for venting of the combustion products since the pressure history will be considerably altered as the shock propagates through these hot gases

  11. 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

  12. 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.

  13. Prospective analysis. Nuclear deterrence in 2030

    International Nuclear Information System (INIS)

    Tertrais, B.

    2006-12-01

    This study is a prospective analysis of the long-term future of nuclear weapons, and particularly the future of French nuclear deterrence after 2015. The selected time period is 2025-2030. The principal objective is to reflect on what the nuclear world might look like during the first part of the 21 st century, beyond the modernization decisions already planned or envisaged, and to draw conclusions for the future of the French deterrent. (author)

  14. Seismic resistant design of a nuclear category I earth dam

    International Nuclear Information System (INIS)

    Vaidya, N.R.; Ries, E.R.; Kissenpfennig, J.F.

    1975-01-01

    An integral part of many nuclear power plants is the ultimate heat sink (UHS); the purpose of which is to retain and deliver a supply of service water to the plant when water from the primary circulating water system is not available. The earth dam described herein is designed to retain the reservoir for the UHS of a nuclear power plant in Southern Europe. The usual pseudo-static analysis is only as good as the estimate for the seismic coefficient used to compute an equivalent horizontal static force on a potential sliding mass. In view of the earth dam considered herein, a more accurate computation of the seismic coefficients is to be made. A two-dimensional dynamic finite element analysis is made to predict the response of the earth dam to a Safe Shutdown Earthquake excitation which is in the form of a time history of accelerations appropriately deconvoluted from the surficial time history and applied at the base of the model. The material properties such as shear modulus and damping are adjusted to be compatible with the level of strain obtained. Thus, non-linear behavior of soil is considered in the analysis and a more realistic response is predicted. Acceleration and stress are determined throughout the dam and are used to compute a seismic coefficient for a pseudo-static stability analysis and the dynamic strength to stress ratios at several points in the body of the dam. The need to design the dam to resist a progressive erosion accident resulting from postulated concentrated leaks is discussed. This may be accomplished by providing a wide, well graded core protected by wide transition cores also heavily compacted

  15. 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.

  16. 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.)

  17. 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

  18. 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.)

  19. Nuclear powerplant design and issues of operation

    International Nuclear Information System (INIS)

    Howard, E.

    1984-01-01

    Various alternatives being looked at by electric utilities management considering additional or replacement nuclear power units are noted. Examples are sizing of units, service life extension of equipment, incremental capacity improvements, load management and pricing, robotics for reduction of occupational exposure, technical advances in water chemistry management and materials, information systems, reactor safety and emergency planning

  20. 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

  1. Nuclear analysis techniques and environmental sciences

    International Nuclear Information System (INIS)

    1997-10-01

    31 theses are collected in this book. It introduced molecular activation analysis micro-PIXE and micro-probe analysis, x-ray fluorescence analysis and accelerator mass spectrometry. The applications about these nuclear analysis techniques are presented and reviewed for environmental sciences

  2. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Selvatici, E.

    1981-01-01

    A study about the safety analysis of nuclear power plant, giving emphasis to how and why to do is presented. The utilization of the safety analysis aiming to perform the licensing requirements is discussed, and an example of the Angra 2 and 3 safety analysis is shown. Some presented tendency of the safety analysis are presented and examples are shown.(E.G.) [pt

  3. Analysis of archaeological pieces with nuclear techniques

    International Nuclear Information System (INIS)

    Tenorio, D.

    2002-01-01

    In this work nuclear techniques such as Neutron Activation Analysis, PIXE, X-ray fluorescence analysis, Metallography, Uranium series, Rutherford Backscattering for using in analysis of archaeological specimens and materials are described. Also some published works and thesis about analysis of different Mexican and Meso american archaeological sites are referred. (Author)

  4. Design Provisions for Withstanding Station Blackout at Nuclear Power Plants

    International Nuclear Information System (INIS)

    2015-08-01

    International operating experience has shown that the loss of off-site power supply concurrent with a turbine trip and unavailability of the standby alternating current power system is a credible event. Lessons learned from the past and recent station blackout events, as well as the analysis of the safety margins performed as part of the ‘stress tests’ conducted on European nuclear power plants in response to the Fukushima Daiichi accident, have identified the station blackout event as a limiting case for most nuclear power plants. The magnitude 9.0 earthquake and consequential tsunami which occurred in Fukushima, Japan, in March 2011, led to a common cause failure of on-site alternating current electrical power supply systems at the Fukushima Daiichi nuclear power plant as well as the off-site power grid. In addition, the resultant flooding caused the loss of direct current power supply, which further exacerbated an already critical situation at the plant. The loss of electrical power resulted in the meltdown of the core in three reactors on the site and severely restricted heat removal from the spent fuel pools for an extended period of time. The plant was left without essential instrumentation and controls, and this made accident management very challenging for the plant operators. The operators attempted to bring and maintain the reactors in a safe state without information on the vital plant parameters until the power supply was eventually restored after several days. Although the Fukushima Daiichi accident progressed well beyond the expected consequences of a station blackout, which is the complete loss of all alternating current power supplies, many of the lessons learned from the accident are valid. A failure of the plant power supply system such as the one that occurred at Fukushima Daiichi represents a design extension condition that requires management with predesigned contingency planning and operator training. The extended loss of all power at a

  5. 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.

  6. 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

  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)

  8. Advanced plant design recommendations from Cook Nuclear Plant experience

    International Nuclear Information System (INIS)

    Zimmerman, W.L.

    1993-01-01

    A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

  9. A world-class design for the Chinese Nuclear Program

    International Nuclear Information System (INIS)

    Ruiz, A.; Valero, E.

    1999-01-01

    Under the name Grupo Nuclear SEPI, the companies BWE, ENDESA, ENSA, ENRESA, ENUSA and INITEC, in a partnership in SEPI and with extensive experience in the nuclear sector, have in the last few years been coordinating technical and commercial actions targeted primarily at the Chinese nuclear market. These actions have been carried out under Westinghouse's leadership, for the purpose of providing a 1000 MWe nuclear Power plant design tailored to the local requirements defined by experts and authorities of the People's Republic of China (project CPWR 1000). (Author)

  10. Nuclear power plant functions: overview, maintenance, design practices, training

    International Nuclear Information System (INIS)

    Gray, J.

    1984-01-01

    The author gives a history of the nuclear industry in the US beginning with the Atoms for Peace Proposal in 1954 and summarizes the nuclear industry's importance in the realm of electric power production today. The primary problems facing the domestic nuclear industry are identified as the lengthening schedules for plant licensing and construction, and the associated uncertainty in plant costs and difficulty in financing, and the erosion of public confidence. Views on technological approaches to the future of nuclear power and the role regulation will play in the future as a fundamental force are discusses in the paper. Also discussed are the importance of standardization of advanced reactor designs and quality assurance

  11. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  12. 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)

  13. 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)

  14. Use of nuclear data sensitivity and uncertainty analysis for the design preparation of the HCLL breeder blanket mock-up experiment for ITER

    International Nuclear Information System (INIS)

    Kodeli, I.

    2007-01-01

    An experiment on a mock-up of the Test Blanket module based on Helium Cooled Lithium Lead (HCLL) concept will be performed in 2007 in the FNG utility in Frascati in order to study neutronics characteristics of the module and the performance of the computational tools in the accurate prediction of the neutron transport. With the objective to prepare and optimise the design of the mock-up in the sense to provide maximum information on the state-of-the-art of the cross section data the mock-up was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross sections, as well as the corresponding uncertainty estimations were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D and the VITAMIN-J/COVA covariance matrix libraries. The cross section reactions with largest contribution to the uncertainty in the calculation of the TPR were identified to be (n,2n) and (n,3n) reactions on plumb. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross sections. (author)

  15. Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

    Full Text Available An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL concept will be performed in 2008 in the Frascati Neutron Generator (FNG in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR, their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package, the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n and (n,3n reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

  16. 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.)

  17. 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.

  18. 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)

  19. Toshiba integrated information system for design of nuclear power plants

    International Nuclear Information System (INIS)

    Abe, Yoko; Kawamura, Hirobumi; Sasaki, Norio; Takasaka, Kiyoshi

    1993-01-01

    TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plants and has been introducing Computer Aided Engineering (CAE). Up to the present, TOSHIBA has been developing computer systems which support each field of design and applying them to the design of nuclear power plants. The new design support system has been developed to integrate each of those systems in order to realize much greater improvement in accuracy and increase of reliability in design using state-of-the-art computer technology

  20. Design of concrete structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

  1. Risk analysis for nuclear power plants

    International Nuclear Information System (INIS)

    Koelzer, W.

    1983-01-01

    The German risk analysis program for nuclear power plants aiming at the man and the environment is presented. An accident consequence model to calculate the radiological impact and the potential health effects is described. (E.G.) [pt

  2. Analysis Of Natural Zeolites For Technical Nuclear

    International Nuclear Information System (INIS)

    Sarria, P.; Desdin, L.; Dominguez, O.

    1999-01-01

    In this article a methodology of elementary analysis of natural zeolites is reported using different technical nuclear (AANR, FRX, MRN and EG). Determines the elementary composition of ours of two Cuban locations. (Author) [es

  3. 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)

  4. 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

  5. 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.)

  6. Research design and statistical analysis

    CERN Document Server

    Myers, Jerome L; Lorch Jr, Robert F

    2013-01-01

    Research Design and Statistical Analysis provides comprehensive coverage of the design principles and statistical concepts necessary to make sense of real data.  The book's goal is to provide a strong conceptual foundation to enable readers to generalize concepts to new research situations.  Emphasis is placed on the underlying logic and assumptions of the analysis and what it tells the researcher, the limitations of the analysis, and the consequences of violating assumptions.  Sampling, design efficiency, and statistical models are emphasized throughout. As per APA recommendations

  7. Nuclear challenges and progress in designing stellarator power plants

    International Nuclear Information System (INIS)

    El-Guebaly, L.

    2007-01-01

    As an alternate to the mainline magnetic fusion tokamaks, the stellarator concept offers a steady state operation without external driven current, eliminating the risk of plasma irruptions. Over the past 2-3 decades, stellarator power plants have been studied in the U.S., Japan, and Europe to enhance the physics and engineering aspects and optimize the design parameters that are subject to numerous constraints. The earlier 1980's studies delivered large stellarators with an average major radius exceeding 20 m. The most recent development of the compact stellarator concept has led to the construction of the National Compact Stellarator Experiment (NCSX) in the U.S. and the 3 years power plant study of ARIES-CS, a compact stellarator with 7.75 m average major radius, approaching that of tokamaks. The ARIES-CS first wall configuration deviates from the standard practice of uniform toroidal shape in order to achieve compactness. Modeling such a complex geometry for 3-D nuclear analysis was a challenging engineering task. A novel approach based on coupling the CAD model with the MCNP Monte Carlo code was developed to model, for the first time ever, the complex stellarator geometry for nuclear assessments. The most important parameter that determines the stellarator size and cost is the minimum distance between the plasma boundary and mid-coil. Accommodating the breeding blanket and necessary shield to protect the superconducting magnet represented another challenging task. An innovative approach utilizing a non-uniform blanket combined with a highly efficient WC shield for this highly constrained area reduced the radial standoff (and machine size and cost) by 25- 30%, which is significant. As stellarators generate more radwaste than tokamaks, managing ARIES-CS active materials during operation and after plant decommissioning was essential for the environmental attractiveness of the machine. The geological disposal option could be replaced with more attractive scenarios

  8. Error analysis of nuclear power plant operator cognitive behavior

    International Nuclear Information System (INIS)

    He Xuhong; Zhao Bingquan; Chen Yulong

    2001-01-01

    Nuclear power plant is a complex human-machine system integrated with many advanced machines, electron devices and automatic controls. It demands operators to have high cognitive ability and correct analysis skill. The author divides operator's cognitive process into five stages to analysis. With this cognitive model, operator's cognitive error is analysed to get the root causes and stages that error happens. The results of the analysis serve as a basis in design of control rooms and training and evaluation of operators

  9. Economic Analysis of Nuclear Energy

    International Nuclear Information System (INIS)

    Lee, Man Ki; Moon, K. H.; Kim, S. S.; Lim, C. Y.; Song, K. D.; Lee, H. M.; Oh, K. B.

    2003-12-01

    This study consists of various issues as follows; electricity price regulation in the liberalized electricity market, establishment of carbon emission limit in national electricity sector, the role of nuclear power as an future energy supply option, the future prospect of CO2 capture and sequestration and current research status of that area in Korea, and Preliminary economic feasibility study of MIP(Medical Isotopes Producer). In the price regulation in the liberalized electricity market, the characteristic of liberalized electricity market in terms of regulation was discussed. The current status and future projection of GHG emission in Korean electricity sector was also investigated. After that, how to set the GHG emission limit in the national electricity sector was discussed. The characteristic of nuclear technology and the research in progress were summarized with the suggestion of the possible new application of nuclear power. The current status and future prospect of the CO2 capture and sequestration research was introduced and current research status of that area in Korea was investigated. Preliminary economic feasibility study of MIP(Medical Isotopes Producer), using liquid nuclear fuel to produce medical isotopes of Mo-99 and Sr-89, was performed

  10. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

  11. 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

  12. 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.)

  13. 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

  14. 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)

  15. Spray pond design for nuclear power plants

    International Nuclear Information System (INIS)

    Codell, R.B.; Asce, A.M.

    1986-01-01

    This paper presents a complex methodology for assessing the performance of spray ponds in ultimate heat sink service at nuclear power plants. A spray pond performance model, developed in the companion paper, is used in conjunction with on-site and off-site meteorological data to predict the highest temperature and greatest 30 day water loss which can reasonable be expected to occur during the lifetime of the plant. The performance model for heat and mass transfer is used to develop an efficient phenomenological model used to scan the long-term meteorological records. Refined estimates of temperature or water loss may then be based on more complicated models if necessary. Short-term onsite data are correlated to the long-term off-site data to formulate correction factors for the difference in location. Cumulative distribution functions for temperature and water loss are determined from the long-term meteorological records to predict the occurrence of these quantities which are less severe that the peak. The methodology is demonstrated using data and parameters from the Palo Verde nuclear plant as an example

  16. Analysis and design of advice

    CERN Document Server

    Jureta, Ivan

    2011-01-01

    This innovative book offers a rigorous approach to the analysis and design of advice in real-world decision situations, in which the advisor must manage with variously imprecise, unclear, incomplete or conflicting qualitative information.

  17. 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.)

  18. 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.)

  19. Thermionic reactor power conditioner design for nuclear electric propulsion.

    Science.gov (United States)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  20. 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)

  1. Conceptual design of nuclear power plants database system

    International Nuclear Information System (INIS)

    Ishikawa, Masaaki; Izumi, Fumio; Sudoh, Takashi.

    1984-03-01

    This report is the result of the joint study on the developments of the nuclear power plants database system. The present conceptual design of the database system, which includes Japanese character processing and image processing, has been made on the data of safety design parameters mainly found in the application documents for reactor construction permit made available to the public. (author)

  2. Modern tornado design of nuclear and other potentially hazardous facilities

    International Nuclear Information System (INIS)

    Stevenson, J.D.; Zhao, Y.

    1996-01-01

    Tornado wind loads and other tornado phenomena, including tornado missiles and differential pressure effects, have not usually been considered in the design of conventional industrial, commercial, or residential facilities in the United States; however, tornado resistance has often become a design requirement for certain hazardous facilities, such as large nuclear power plants and nuclear materials and waste storage facilities, as well as large liquefied natural gas storage facilities. This article provides a review of current procedures for the design of hazardous industrial facilities to resist tornado effects. 23 refs., 19 figs., 13 tabs

  3. Engine cycle design considerations for nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Collins, J.T.

    1993-01-01

    A top-level study was performed which addresses nuclear thermal propulsion system engine cycle options and their applicability to support future Space Exploration Initiative manned lunar and Mars missions. Technical and development issues associated with expander, gas generator, and bleed cycle near-term, solid core nuclear thermal propulsion engines are identified and examined. In addition to performance and weight the influence of the engine cycle type on key design selection parameters such as design complexity, reliability, development time, and cost are discussed. Representative engine designs are presented and compared. Their applicability and performance impact on typical near-term lunar and Mars missions are shown

  4. Chemical analysis by nuclear methods. v. 2

    International Nuclear Information System (INIS)

    Alfassi, Z.B.

    1998-01-01

    'Chemical analysis by Nuclear Methods' is an effort of some renowned authors in field of nuclear chemistry and radiochemistry which is compiled by Alfassi, Z.B. and translated into Farsi version collected in two volumes. The second volume consists of the following chapters: Detecting ion recoil scattering and elastic scattering are dealt in the eleventh chapter, the twelfth chapter is devoted to nuclear reaction analysis using charged particles, X-ray emission is discussed at thirteenth chapter, the fourteenth chapter is about using ion microprobes, X-ray fluorescence analysis is discussed in the fifteenth chapter, alpha, beta and gamma ray scattering in chemical analysis are dealt in chapter sixteen, Moessbauer spectroscopy and positron annihilation are discussed in chapter seventeen and eighteen; The last two chapters are about isotope dilution analysis and radioimmunoassay

  5. 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

  6. 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)

  7. Planning and architectural safety considerations in designing nuclear power plants

    International Nuclear Information System (INIS)

    Konsowa, Ahmed A.

    2009-01-01

    To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

  8. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    IAS Admin

    technologies which produce 5780 MW of electric power. Reactors are .... The fuel pin is the elementary entity of the NR. We shall arrive at the ..... Design is given short shrift in physics education both in high school and in college. The simple ...

  9. 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.)

  10. Incorporating design for decommissioning into the layout of nuclear facilities

    International Nuclear Information System (INIS)

    Collum, B.; Druart, A.

    2008-01-01

    Design for Decommissioning (DfD) is the design of nuclear facilities in a manner that facilitates ultimate decommissioning in as safe, technically efficient and cost effective way as possible. Strictly speaking, (DfD) should need minimal introduction and this paper should ideally be aimed at discussing the finer points of some improvement to a practice that is already widely embedded throughout the nuclear industry. The reality though is quite different. As an industry, we all know what DfD is and indeed we do incorporate it into our designs. However, application is at best patchy and there is little evidence of applying it to the level that will be advocated here. When applied at its highest level, DfD is all about truly designing nuclear facilities with their whole life cycle in mind, such that the decommissioning phase is an integral part of the design of a facility from the very first day. In this way, when a facility comes to the end of its operational life, it can move smoothly to Post Operational Clean Out (POCO) and then through the various phases of decommissioning. Demonstrating from the start that the nuclear industry addresses the challenges posed by decommissioning will help it to gain support from the regulators and the general public for proposals to build new nuclear generating capacity. (author)

  11. 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

  12. Nuclear data sets for reactor design calculations - approved 1975

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  13. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  14. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  15. 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

  16. 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)

  17. Defense against nuclear weapons: a decision analysis

    International Nuclear Information System (INIS)

    Orient, J.M.

    1985-01-01

    Response to the public health threat posed by nuclear weapons is a medical imperative. The United States, in contrast to other nations, has chosen a course that assures maximal casualties in the event of a nuclear attack, on the theory that prevention of the attack is incompatible with preventive measures against its consequences, such as blast injuries and radiation sickness. A decision analysis approach clarifies the risks and benefits of a change to a strategy of preparedness

  18. The nuclear analysis program at MURR

    International Nuclear Information System (INIS)

    Glascock, M.D.

    1993-01-01

    The University of Missouri-Columbia (MU) has continually upgraded research facilities and programs at the MU research reactor (MURR) throughout its 26-yr history. The Nuclear Analysis Program (NAP) area has participated in these upgrades over the years. As one of the largest activation analysis laboratories on a university campus, the activities of the NAP are broadly representative of the diversity of applications for activation analysis and related nuclear science. This paper describes the MURR's NAP and several of the research, education, and service projects in which the laboratory is currently engaged

  19. 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)

  20. Stress Analysis for Mobile Hot Cell Design

    International Nuclear Information System (INIS)

    Muhammad Hannan Bahrin; Anwar Abdul Rahman; Mohd Arif Hamzah

    2015-01-01

    Prototype and Plant Development Centre (PDC) is developing a Mobile Hot Cell (MHC) to handle and manage Spent High Activity Radioactive Sources (SHARS), such as teletherapy heads and dry irradiators. At present, there are two units of MHC in the world, one in South Africa and the other one in China. Malaysian Mobile MHC is developed by Malaysian Nuclear Agency with the assistance of IAEA expert, based on the design of South Africa and China, but with improved features. Stress analysis has been performed on the design to fulfill the safety requirement in MHC operation. This paper discusses the loading effect analysis from the radiation shielding materials to the MHC wall structure, roof supporting column and window structure. (author)

  1. 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

  2. Nuclear class 1 piping stress analysis

    International Nuclear Information System (INIS)

    Lucas, J.C.R.; Maneschy, J.E.; Mariano, L.A.; Tamura, M.

    1981-01-01

    A nuclear class 1 piping stress analysis, according to the ASME code, is presented. The TRHEAT computer code has been used to determine the piping wall thermal gradient. The Nupipe computer code was employed for the piping stress analysis. Computer results were compared with the allowable criteria from the ASME code. (Author) [pt

  3. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Owen, P.S.; Parker, M.B.; Omberg, R.P.

    1979-05-01

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U 3 O 8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  4. 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

  5. Design of nuclear cells with re linking of trajectories

    International Nuclear Information System (INIS)

    Castillo, A.; Campos S, Y.; Ortiz S, J.J.; Montes, J.L.; Perrusquia, R.; Hernandez, J.L.; Torres, M.

    2006-01-01

    Presently work the results obtained with the Ohtli-RT system obtained when implementing the combinatory optimization technique well-known as Trajectories re linking or Path Re linking in English. The problem to solve is the radial design of nuclear fuel, taking like base nuclear fuel assembles for boiling water reactors (BWR Boiling Water Reactor by its initials in English). To evaluate the objective function used in the system the code in two dimensions Heliums 1.5 was used, which calculates the cross sections of the proposed design. The parameters that were considered for the evaluation of the objective function are the Power peak factor of the bar that generates bigger power in the cell and the Infinite Multiplication Factor. To prove the system its were used assembles 10x10 with 2 water channels. The obtained radial designs of nuclear fuel fulfilled the restrictions imposed to the considered limits, with regard to the involved parameters. (Author)

  6. 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

  7. Reliability Analysis Techniques for Communication Networks in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lim, T. J.; Jang, S. C.; Kang, H. G.; Kim, M. C.; Eom, H. S.; Lee, H. J.

    2006-09-01

    The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for nuclear power plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of this study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

  8. 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)

  9. 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)

  10. 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.)

  11. Cycle design flexibility for the Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Beu, T.D.; Keys, T.A.; Gardner, D.R.

    1986-01-01

    Fuel cycle length flexibility is being included in the cycle designs of the Browns Ferry Nuclear Plant owned and operated by the Tennessee Valley Authority. Large end-of-cycle exposure windows are accommodated by adjusting the length of end-of-cycle coastdowns. Constraints on cycle designs are being lessened by application of innovative core loading strategies and through implementation of advanced fuel designs. Changes in design bases are evaluated relatively quickly and factored into cycle designs in order to maintain or improve performance

  12. 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.)

  13. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  14. Nuclear design of a very-low-activation fusion reactor

    International Nuclear Information System (INIS)

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design

  15. AP1000, a nuclear central of advanced design

    International Nuclear Information System (INIS)

    Hernandez M, N.; Viais J, J.

    2005-01-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  16. Optimal design of condenser volume in nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Jing; Yan Changqi; Wang Jianjun

    2011-01-01

    The condenser is an important component in the nuclear power plant,whose dimension will influence the economy and the arrangement of the nuclear power plant.In this paper, the calculation model was established according to the design experience. The corresponding codes were also developed. The sensitivity of design parameters which influence the condenser Janume was analyzed. The present optimal design of the condenser, aiming at the volume minimization, was carried out with the self-developed complex-genetic algorithm. The results show that the reference condenser design is far from the best scheme. In addition, the results also verify the feasibility of the complex-genetic algorithm. Furthermore, the results of this paper can provide reference for the design of the condenser. (authors)

  17. Planning, design and technological criteria of conventional and nuclear shelters

    International Nuclear Information System (INIS)

    Sadoon, A.S.

    1989-01-01

    The thesis aims to establish a special criteria for building the shelters in two types. The conventional and nuclear, in respect to planning design and technological aspects, and finally establishing a special reference of planning, design and technology for Iraq which can be used when planning or designing a conventional or nuclear shelter. The thesis included four chapters, the first chapter included definition of shelters, and explanation of the effects of all types of weapons on buildings, and the second chapter included definition of planning and design concepts of shelters in its two types and analytical studies for international examples. The third chapter covered definition for technologies of structural, mechanical, electrical and sanitary systems. The fourth chapter included details of a case study in order to approach the results of research which included the conclusions, recommendations, criteria and prospects of planning design and technological aspects. 51 tabs.; 180 figs.; 32 refs.; 15 apps

  18. 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)

  19. System design for safe robotic handling of nuclear materials

    International Nuclear Information System (INIS)

    Drotning, W.; Wapman, W.; Fahrenholtz, J.; Kimberly, H.; Kuhlmann, J.

    1996-01-01

    Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive nuclear materials. These systems will reduce the occupational radiation exposure to workers by automating operations which are currently performed manually. Because the robotic systems will handle material that is both hazardous and valuable, the safety of the operations is of utmost importance; assurance must be given that personnel will not be harmed and that the materials and environment will be protected. These safety requirements are met by designing safety features into the system using a layered approach. Several levels of mechanical, electrical and software safety prevent unsafe conditions from generating a hazard, and bring the system to a safe state should an unexpected situation arise. The system safety features include the use of industrial robot standards, commercial robot systems, commercial and custom tooling, mechanical safety interlocks, advanced sensor systems, control and configuration checks, and redundant control schemes. The effectiveness of the safety features in satisfying the safety requirements is verified using a Failure Modes and Effects Analysis. This technique can point out areas of weakness in the safety design as well as areas where unnecessary redundancy may reduce the system reliability

  20. The design study of the JT-60SU device. No.8. Nuclear shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

    Miya, Naoyuki; Kikuchi, Mitsuru; Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-03-01

    Results of nuclear shielding design study and safety analysis for the steady-state tokamak device JT-60SU are described. D-T operation (option) for two years is adopted in addition to ten years operation using deuterium. Design work has been done in accordance with general laws for radioisotopes handling in Japan as a guideline of safety evaluation, which is applied to the operation of present JT-60U device. Optimization of the shielding design for the device structure including vacuum vessel has been presented to meet with allowable limits of biological shielding determined in advance. It is shown that JT-60SU can be operated safely in the present JT-60 experimental building. It is planed to use 100g/year of tritium in D-T operation phase. A concept of multiple -barrier system is applied to the facility design to prevent propagation of tritium, in which the torus hall and the tritium removal room provide the tertiary confinement. From the design of atmosphere detritiation system for accidental tritium release, it is shown that tritium concentration level can be reduced to the allowable level after two weeks with reasonable compact size components. Safety assessment related to activation of coolant/air, and atmospheric tritium effluents are discussed. (author)

  1. 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

  2. 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

  3. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  4. 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

  5. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  6. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan's effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  7. 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

  8. 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

  9. Network Analysis, Architecture, and Design

    CERN Document Server

    McCabe, James D

    2007-01-01

    Traditionally, networking has had little or no basis in analysis or architectural development, with designers relying on technologies they are most familiar with or being influenced by vendors or consultants. However, the landscape of networking has changed so that network services have now become one of the most important factors to the success of many third generation networks. It has become an important feature of the designer's job to define the problems that exist in his network, choose and analyze several optimization parameters during the analysis process, and then prioritize and evalua

  10. A statistical approach to nuclear fuel design and performance

    Science.gov (United States)

    Cunning, Travis Andrew

    As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

  11. Exome Array Analysis of Nuclear Lens Opacity.

    Science.gov (United States)

    Loomis, Stephanie J; Klein, Alison P; Lee, Kristine E; Chen, Fei; Bomotti, Samantha; Truitt, Barbara; Iyengar, Sudha K; Klein, Ronald; Klein, Barbara E K; Duggal, Priya

    2018-06-01

    Nuclear cataract is the most common subtype of age-related cataract, the leading cause of blindness worldwide. It results from advanced nuclear sclerosis, or opacity in the center of the optic lens, and is affected by both genetic and environmental risk factors, including smoking. We sought to understand the genetic factors associated with nuclear sclerosis through interrogation of rare and low frequency coding variants using exome array data. We analyzed Illumina Human Exome Array data for 1,488 participants of European ancestry in the Beaver Dam Eye Study who were without cataract surgery for association with nuclear sclerosis grade, controlling for age and sex. We performed single-variant regression analysis for 32,138 variants with minor allele frequency (MAF) ≥0.003. In addition, gene-based analysis of 11,844 genes containing at least two variants with MAF nuclear sclerosis, the possible association with the RNF149 gene highlights a potential candidate gene for future studies that aim to understand the genetic architecture of nuclear sclerosis.

  12. Probabilistic Design and Analysis Framework

    Science.gov (United States)

    Strack, William C.; Nagpal, Vinod K.

    2010-01-01

    PRODAF is a software package designed to aid analysts and designers in conducting probabilistic analysis of components and systems. PRODAF can integrate multiple analysis programs to ease the tedious process of conducting a complex analysis process that requires the use of multiple software packages. The work uses a commercial finite element analysis (FEA) program with modules from NESSUS to conduct a probabilistic analysis of a hypothetical turbine blade, disk, and shaft model. PRODAF applies the response surface method, at the component level, and extrapolates the component-level responses to the system level. Hypothetical components of a gas turbine engine are first deterministically modeled using FEA. Variations in selected geometrical dimensions and loading conditions are analyzed to determine the effects of the stress state within each component. Geometric variations include the cord length and height for the blade, inner radius, outer radius, and thickness, which are varied for the disk. Probabilistic analysis is carried out using developing software packages like System Uncertainty Analysis (SUA) and PRODAF. PRODAF was used with a commercial deterministic FEA program in conjunction with modules from the probabilistic analysis program, NESTEM, to perturb loads and geometries to provide a reliability and sensitivity analysis. PRODAF simplified the handling of data among the various programs involved, and will work with many commercial and opensource deterministic programs, probabilistic programs, or modules.

  13. Analysis of renal nuclear medicine images

    International Nuclear Information System (INIS)

    Jose, R.M.J.

    2000-01-01

    Nuclear medicine imaging of the renal system involves producing time-sequential images showing the distribution of a radiopharmaceutical in the renal system. Producing numerical and graphical data from nuclear medicine studies requires defining regions of interest (ROIs) around various organs within the field of view, such as the left kidney, right kidney and bladder. Automating this process has several advantages: a saving of a clinician's time; enhanced objectivity and reproducibility. This thesis describes the design, implementation and assessment of an automatic ROI generation system. The performance of the system described in this work is assessed by comparing the results to those obtained using manual techniques. Since nuclear medicine images are inherently noisy, the sequence of images is reconstructed using the first few components of a principal components analysis in order to reduce the noise in the images. An image of the summed reconstructed sequence is then formed. This summed image is segmented by using an edge co-occurrence matrix as a feature space for simultaneously classifying regions and locating boundaries. Two methods for assigning the regions of a segmented image to organ class labels are assessed. The first method is based on using Dempster-Shafer theory to combine uncertain evidence from several sources into a single evidence; the second method makes use of a neural network classifier. The use of each technique in classifying the regions of a segmented image are assessed in separate experiments using 40 real patient-studies. A comparative assessment of the two techniques shows that the neural network produces more accurate region labels for the kidneys. The optimum neural system is determined experimentally. Results indicate that combining temporal and spatial information with a priori clinical knowledge produces reasonable ROIs. Consistency in the neural network assignment of regions is enhanced by taking account of the contextual

  14. 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.

  15. To MARS and Beyond with Nuclear Power - Design Concept of Korea Advanced Nuclear Thermal Engine Rocket

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Chang, Soon Heung

    2013-01-01

    The President Park of ROK has also expressed support for space program promotion, praising the success of NARO as evidence of a positive outlook. These events hint a strong signal that ROK's space program will be accelerated by the national eager desire. In this national eager desire for space program, the policymakers and the aerospace engineers need to pay attention to the advanced nuclear technology of ROK that is set to a major world nuclear energy country, even exporting the technology. The space nuclear application is a very much attractive option because its energy density is the most enormous among available energy sources in space. This paper presents the design concept of Korea Advanced Nuclear Thermal Engine Rocket (KANuTER) that is one of the advanced nuclear thermal rocket engine developing in Korea Advanced Institute of Science and Technology (KAIST) for space application. Solar system exploration relying on CRs suffers from long trip time and high cost. In this regard, nuclear propulsion is a very attractive option for that because of higher performance and already demonstrated technology. Although ROK was a late entrant into elite global space club, its prospect as a space racer is very bright because of the national eager desire and its advanced technology. Especially it is greatly meaningful that ROK has potential capability to launch its nuclear technology into space as a global nuclear energy leader and a soaring space adventurer. In this regard, KANuTER will be a kind of bridgehead for Korean space nuclear application

  16. To MARS and Beyond with Nuclear Power - Design Concept of Korea Advanced Nuclear Thermal Engine Rocket

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

    The President Park of ROK has also expressed support for space program promotion, praising the success of NARO as evidence of a positive outlook. These events hint a strong signal that ROK's space program will be accelerated by the national eager desire. In this national eager desire for space program, the policymakers and the aerospace engineers need to pay attention to the advanced nuclear technology of ROK that is set to a major world nuclear energy country, even exporting the technology. The space nuclear application is a very much attractive option because its energy density is the most enormous among available energy sources in space. This paper presents the design concept of Korea Advanced Nuclear Thermal Engine Rocket (KANuTER) that is one of the advanced nuclear thermal rocket engine developing in Korea Advanced Institute of Science and Technology (KAIST) for space application. Solar system exploration relying on CRs suffers from long trip time and high cost. In this regard, nuclear propulsion is a very attractive option for that because of higher performance and already demonstrated technology. Although ROK was a late entrant into elite global space club, its prospect as a space racer is very bright because of the national eager desire and its advanced technology. Especially it is greatly meaningful that ROK has potential capability to launch its nuclear technology into space as a global nuclear energy leader and a soaring space adventurer. In this regard, KANuTER will be a kind of bridgehead for Korean space nuclear application.

  17. 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)

  18. 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.

  19. Design of Radioactive Waste Management Systems at Nuclear Power Plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide is addressed to the administrative and technical authorities and specialists dealing with the design, construction and operation of nuclear power plants, and in particular waste management facilities at nuclear power plants. This Guide has been prepared as part of the IAEA Waste Handling, Treatment and Storage programme. It is a follow-up document to the Code of Practice on Management of Radioactive Wastes from Nuclear Power Plants published in 1985 in the IAEA Safety Standards, Safety Series No. 69, in which basic principles for management of radioactive wastes at nuclear power plants are set out. The IAEA has established wide ranging programmes to provide Member States with guidance on different aspects of safety and technology related to thermal neutron power reactors and associated nuclear fuel cycle operations, including those for management of radioactive wastes. There are many IAEA publications related to various technical and safety aspects of different nuclear energy applications. All these publications are issued by the Agency for the use of Member States in connection with their own nuclear technological safety requirements. They are based on national experience contributed by experts from different countries and relate to common features in approaches to the problems discussed. However, the final decision and legal responsibility in any regulatory procedure always rest with the Member State. This particular Guide aims to provide general and detailed principles for the design of waste management facilities at nuclear power plants. It emphasizes what and how specific safety requirements for the management of radioactive wastes from nuclear power plants can be met in the design and construction stage. The safety requirements for operation of such facilities will be considered in the Agency's next Safety Series publication, Safety Guide 50-SG-011, Operational Management for Radioactive Effluents and Wastes Arising in Nuclear Power Plants

  20. 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

  1. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  2. Design of nuclear power generation plants adopting model engineering method

    International Nuclear Information System (INIS)

    Waki, Masato

    1983-01-01

    The utilization of model engineering as the method of design has begun about ten years ago in nuclear power generation plants. By this method, the result of design can be confirmed three-dimensionally before actual production, and it is the quick and sure method to meet the various needs in design promptly. The adoption of models aims mainly at the improvement of the quality of design since the high safety is required for nuclear power plants in spite of the complex structure. The layout of nuclear power plants and piping design require the model engineering to arrange rationally enormous quantity of things in a limited period. As the method of model engineering, there are the use of check models and of design models, and recently, the latter method has been mainly taken. The procedure of manufacturing models and engineering is explained. After model engineering has been completed, the model information must be expressed in drawings, and the automation of this process has been attempted by various methods. The computer processing of design is in progress, and its role is explained (CAD system). (Kako, I.)

  3. 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.)

  4. 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)

  5. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    Bambang Galung Susanto

    2007-01-01

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  6. 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

  7. An assessment of design control practices and design reconstitution programs in the nuclear power industry

    International Nuclear Information System (INIS)

    Imbro, E.V.

    1991-02-01

    The US Nuclear Regulatory Commission (NRC) and the utilities have identified shortcomings involving the maintenance of well-defined design bases and the availability of the necessary supporting design documentation. Many utilities have embarked on design-document reconstitution programs although there has been no clear consensus regarding what information should be included in design-bases documents, what is the minimum set of necessary design documents to support the design bases, or how missing or deficient design documentation should be handled. The NRC initiated a survey to ascertain the status of design control programs within the industry and the approaches to design-bases documentation used by some utilities. The survey scope included six utilities and one nuclear steam supply system vendor. Conclusions and observations resulting from the survey assessments are provided so that utilities and the NRC can consider actions to improve these programs. 12 refs

  8. Promising design options for the encapsulated nuclear heat source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

    2001-07-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  9. Promising design options for the encapsulated nuclear heat source reactor

    International Nuclear Information System (INIS)

    Conway, L.; Carelli, M.D.; Dzodzo, M.; Hossain, Q.; Brown, N.W.; Wade, D.C.; Sienick, J.J.; Greenspan, E.; Kastenberg, W.E.; Saphier, D.

    2001-01-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  10. Integrating design and purchasing [in nuclear engineering] with Ingecad

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Ingecad was developed by the Ingevision division of Framatome to overcome deficiencies in traditional computer-aided design. It was developed for nuclear power project engineering around the principle of the shared management of a common database, thus making it possible to integrate several engineering disciplines. The multiuser database is managed and accessed by the different application softwares, corresponding to particular aspects of the engineering task: electrical and process control schematics; plant piping design; pressurized equipment design etc. The use of a common database ensures coherence between the different engineering disciplines, particularly between the process engineering, the plant layout design, the piping, and the instrumentation and control engineering. (author)

  11. Nuclear power control system design using genetic algorithm

    International Nuclear Information System (INIS)

    Lee, Yoon Joon; Cho, Kyung Ho

    1996-01-01

    The genetic algorithm(GA) is applied to the design of the nuclear power control system. The reactor control system model is described in the LQR configuration. The LQR system order is increased to make the tracking system. The key parameters of the design are weighting matrices, and these are usually determined through numerous simulations in the conventional design. To determine the more objective and optimal weightings, the improved GA is applied. The results show that the weightings determined by the GA yield the better system responses than those obtained by the conventional design method

  12. Heuristic decision model for intelligent nuclear power systems design

    International Nuclear Information System (INIS)

    Nassersharif, B.; Portal, M.G.; Gaeta, M.J.

    1989-01-01

    The objective of this project was to investigate intelligent nuclear power systems design. A theoretical model of the design process has been developed. A fundamental process in this model is the heuristic decision making for design (i.e., selection of methods, components, materials, etc.). Rule-based expert systems do not provide the completeness that is necessary to generate good design. A new method, based on the fuzzy set theory, has been developed and is presented here. A feedwater system knowledge base (KB) was developed for a prototype software experiment to benchmark the theory

  13. French administrative practice and design codes for nuclear vessels

    International Nuclear Information System (INIS)

    Roche, R.L.

    1987-07-01

    French regulations on boilers and pressure vessels have prevailed for a very long time, the first measure having been promulgated on 29 October 1823. Restraining the attention to nuclear pressure vessels it must be pointed out regulations and enforcement by public authorities are more stringent than they are for conventional pressure vessels. The first part of this paper will be devoted to regulations with a special attention to the decree of 26 February 1974 and to the practice of public authorities in this field with special attention given to the Bureau de Controle de la Construction Nucleaire (BCCN = Bureau of Inspection of Nuclear Design and Manufacturing). The second part of this paper will deal with the French construction codes for nuclear components RCC-M (water reactors) and RCC-MR (elevated temperature design)

  14. 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

  15. 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)

  16. Framework for Integrating Safety, Operations, Security, and Safeguards in the Design and Operation of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Darby, John L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Horak, Karl Emanuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaChance, Jeffrey L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Tolk, Keith Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Whitehead, Donnie Wayne [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2007-10-01

    The US is currently on the brink of a nuclear renaissance that will result in near-term construction of new nuclear power plants. In addition, the Department of Energy’s (DOE) ambitious new Global Nuclear Energy Partnership (GNEP) program includes facilities for reprocessing spent nuclear fuel and reactors for transmuting safeguards material. The use of nuclear power and material has inherent safety, security, and safeguards (SSS) concerns that can impact the operation of the facilities. Recent concern over terrorist attacks and nuclear proliferation led to an increased emphasis on security and safeguard issues as well as the more traditional safety emphasis. To meet both domestic and international requirements, nuclear facilities include specific SSS measures that are identified and evaluated through the use of detailed analysis techniques. In the past, these individual assessments have not been integrated, which led to inefficient and costly design and operational requirements. This report provides a framework for a new paradigm where safety, operations, security, and safeguards (SOSS) are integrated into the design and operation of a new facility to decrease cost and increase effectiveness. Although the focus of this framework is on new nuclear facilities, most of the concepts could be applied to any new, high-risk facility.

  17. Bulk analysis using nuclear techniques

    International Nuclear Information System (INIS)

    Borsaru, M.; Holmes, R.J.; Mathew, P.J.

    1983-01-01

    Bulk analysis techniques developed for the mining industry are reviewed. Using penetrating neutron and #betta#-radiations, measurements are obtained directly from a large volume of sample (3-30 kg) #betta#-techniques were used to determine the grade of iron ore and to detect shale on conveyor belts. Thermal neutron irradiation was developed for the simultaneous determination of iron and aluminium in iron ore on a conveyor belt. Thermal-neutron activation analysis includes the determination of alumina in bauxite, and manganese and alumina in manganese ore. Fast neutron activation analysis is used to determine silicon in iron ores, and alumina and silica in bauxite. Fast and thermal neutron activation has been used to determine the soil in shredded sugar cane. (U.K.)

  18. Nuclear plant analyzer development and analysis applications

    International Nuclear Information System (INIS)

    Laats, E.T.

    1984-10-01

    The Nuclear Plant Analyzer (NPA) is being developed as the US Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

  19. Finite Element Analysis and Design of Experiments in Engineering Design

    OpenAIRE

    Eriksson, Martin

    1999-01-01

    Projects with the objective of introducing Finite Element Analysis (FEA) into the early phases of the design process have previously been carried out at the Department of Machine Design, see e.g. the Doctoral thesis by Burman [13]. These works clearly highlight the usefulness of introducing design analysis early in the design process. According to Bjärnemo and Burman [10] the most significant advantage of applying design analysis early in the design process was the shift from verification to ...

  20. Civil engineering in nuclear power stations: design of the turbine building and nuclear auxiliary building

    International Nuclear Information System (INIS)

    Lacroix, R.

    1985-01-01

    After enumerating the specific features of civil engineering in nuclear power stations. One goes on to examine the principal deliberations undertaken with the aim of optimising projects for transition from the P4 to P'4 and then N4 generations of nuclear power stations. The courses of action decided with respect to the design of the machine room and auxiliary equipment building are described [fr