WorldWideScience

Sample records for reactor analysis methodology

  1. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  2. Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Amos, C.N.; Wagner, K.C.; Bradley, D.R. (Science Applications International Corp., Albuquerque, NM (United States))

    1992-01-01

    A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained.

  3. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  4. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  5. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    Science.gov (United States)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  6. A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

    2016-11-01

    In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

  7. A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

    2016-06-26

    GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.

  8. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

    2015-06-28

    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  9. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  10. Sargent-IV Project. Development of new methodologies for safety analysis of Generation IV reactors; Proyecto SARGEB-IV. Desarrollo de nuevas metodologias de analisis de seguridad para reactores de Generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Gallego, E.; Jimenez, G.

    2013-07-01

    The main result of this paper is the proposal for the addition of new ingredients in the safety analysis methodologies for Generation-IV reactors that integrates the features of probabilistic safety analysis within deterministic. This ensures a higher degree of integration between the classical deterministic and probabilistic methodologies.

  11. Application of FORSS sensitivity and uncertainty methodology to fast reactor benchmark analysis

    Energy Technology Data Exchange (ETDEWEB)

    Weisbin, C.R.; Marable, J.H.; Lucius, J.L.; Oblow, E.M.; Mynatt, F.R.; Peelle, R.W.; Perey, F.G.

    1976-12-01

    FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions, and associated uncertainties. This paper presents the theory and code description as well as the first results of applying FORSS to fast reactor benchmarks. Specifically, for various assemblies and reactor performance parameters, the nuclear data sensitivities were computed by nuclide, reaction type, and energy. Comprehensive libraries of energy-dependent coefficients have been developed in a computer retrievable format and released for distribution by RSIC and NNCSC. Uncertainties induced by nuclear data were quantified using preliminary, energy-dependent relative covariance matrices evaluated with ENDF/B-IV expectation values and processed for /sup 238/U(n,f), /sup 238/U(n,..gamma..), /sup 239/Pu(n,f), and /sup 239/Pu(..nu..). Nuclear data accuracy requirements to meet specified performance criteria at minimum experimental cost were determined.

  12. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  13. Analysis of offsite dose calculation methodology for a nuclear power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moser, Donna Smith [Univ. of North Carolina, Chapel Hill, NC (United States)

    1995-01-01

    This technical study reviews the methodology for calculating offsite dose estimates as described in the offsite dose calculation manual (ODCM) for Pennsylvania Power and Light - Susquehanna Steam Electric Station (SSES). An evaluation of the SSES ODCM dose assessment methodology indicates that it conforms with methodology accepted by the US Nuclear Regulatory Commission (NRC). Using 1993 SSES effluent data, dose estimates are calculated according to SSES ODCM methodology and compared to the dose estimates calculated according to SSES ODCM and the computer model used to produce the reported 1993 dose estimates. The 1993 SSES dose estimates are based on the axioms of Publication 2 of the International Commission of Radiological Protection (ICRP). SSES Dose estimates based on the axioms of ICRP Publication 26 and 30 reveal the total body estimates to be the most affected.

  14. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  15. Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''

    Energy Technology Data Exchange (ETDEWEB)

    Dmitriy Y. Anistratov; Marvin L. Adams; Todd S. Palmer; Kord S. Smith; Kevin Clarno; Hikaru Hiruta; Razvan Nes

    2003-08-04

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.

  16. Prometheus Reactor I&C Software Development Methodology, for Action

    Energy Technology Data Exchange (ETDEWEB)

    T. Hamilton

    2005-07-30

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

  17. Risk-assessment methodology for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K. O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed.

  18. Automated scoping methodology for liquid metal natural circulation small reactor

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hyung M.; Suh, Kune Y., E-mail: kysuh@snu.ac.kr

    2014-07-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  19. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

  20. Dynamic analysis of process reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shadle, L.J.; Lawson, L.O.; Noel, S.D.

    1995-06-01

    The approach and methodology of conducting a dynamic analysis is presented in this poster session in order to describe how this type of analysis can be used to evaluate the operation and control of process reactors. Dynamic analysis of the PyGas{trademark} gasification process is used to illustrate the utility of this approach. PyGas{trademark} is the gasifier being developed for the Gasification Product Improvement Facility (GPIF) by Jacobs-Siffine Engineering and Riley Stoker. In the first step of the analysis, process models are used to calculate the steady-state conditions and associated sensitivities for the process. For the PyGas{trademark} gasifier, the process models are non-linear mechanistic models of the jetting fluidized-bed pyrolyzer and the fixed-bed gasifier. These process sensitivities are key input, in the form of gain parameters or transfer functions, to the dynamic engineering models.

  1. Coal conversion processes and analysis methodologies for synthetic fuels production. [technology assessment and economic analysis of reactor design for coal gasification

    Science.gov (United States)

    1979-01-01

    Information to identify viable coal gasification and utilization technologies is presented. Analysis capabilities required to support design and implementation of coal based synthetic fuels complexes are identified. The potential market in the Southeast United States for coal based synthetic fuels is investigated. A requirements analysis to identify the types of modeling and analysis capabilities required to conduct and monitor coal gasification project designs is discussed. Models and methodologies to satisfy these requirements are identified and evaluated, and recommendations are developed. Requirements for development of technology and data needed to improve gasification feasibility and economies are examined.

  2. Regional Shelter Analysis Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Dillon, Michael B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dennison, Deborah [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kane, Jave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Walker, Hoyt [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-01

    The fallout from a nuclear explosion has the potential to injure or kill 100,000 or more people through exposure to external gamma (fallout) radiation. Existing buildings can reduce radiation exposure by placing material between fallout particles and exposed people. Lawrence Livermore National Laboratory was tasked with developing an operationally feasible methodology that could improve fallout casualty estimates. The methodology, called a Regional Shelter Analysis, combines the fallout protection that existing buildings provide civilian populations with the distribution of people in various locations. The Regional Shelter Analysis method allows the consideration of (a) multiple building types and locations within buildings, (b) country specific estimates, (c) population posture (e.g., unwarned vs. minimally warned), and (d) the time of day (e.g., night vs. day). The protection estimates can be combined with fallout predictions (or measurements) to (a) provide a more accurate assessment of exposure and injury and (b) evaluate the effectiveness of various casualty mitigation strategies. This report describes the Regional Shelter Analysis methodology, highlights key operational aspects (including demonstrating that the methodology is compatible with current tools), illustrates how to implement the methodology, and provides suggestions for future work.

  3. A Nuclear Reactor Transient Methodology Based on Discrete Ordinates Method

    Directory of Open Access Journals (Sweden)

    Shun Zhang

    2014-01-01

    Full Text Available With the rapid development of nuclear power industry, simulating and analyzing the reactor transient are of great significance for the nuclear safety. The traditional diffusion theory is not suitable for small volume or strong absorption problem. In this paper, we have studied the application of discrete ordinates method in the numerical solution of space-time kinetics equation. The fully implicit time integration was applied and the precursor equations were solved by analytical method. In order to improve efficiency of the transport theory, we also adopted some advanced acceleration methods. Numerical results of the TWIGL benchmark problem presented demonstrate the accuracy and efficiency of this methodology.

  4. A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

    2016-06-26

    GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.

  5. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

    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.

  6. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  7. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  8. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  9. Methodology for the integral comparison of nuclear reactors: selecting a reactor for Mexico; Metodologia para la comparacion integral de reactores nucleares: seleccion de un reactor para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2006-07-01

    In this work it was built a methodology to compare nuclear reactors of third generation that can be contemplated for future electric planning in Mexico. This methodology understands the selection of the reactors to evaluate eliminating the reactors that still are not sufficiently mature at this time of the study. A general description of each reactor together with their main ones characteristic is made. It was carried out a study for to select the group of parameters that can serve as evaluation indicators, which are the characteristics of the reactors with specific values for each considered technology, and it was elaborated an evaluation matrix indicators including the reactors in the columns and those indicators in the lines. Since that none reactor is the best in all the indicators were necessary to use a methodology for multi criteria taking decisions that we are presented. It was used the 'Fuzzy Logic' technique, the which is based in those denominated diffuse groups and in a system of diffuse inference based on heuristic rules in the way 'If Then consequence> ', where the linguistic values of the condition and of the consequence is defined by diffuse groups, it is as well as the rules always they transform a diffuse group into another. Later on they combine all the diffuse outputs to create a single output and an inverse transformation is made that it transfers the output from the diffuse domain to the real one. They were carried out two studies one with the entirety of the indicators and another in which the indicators were classified in three approaches: the first one refers to indicators related with the planning of the plants inside the context of the general electric sector, the second approach includes indicators related with the characteristics of the fuel and the third it considers indicators related with the acting of the plant in safety and environmental impact. This second study allowed us to know the qualities of

  10. Methodology of Supervision by Analysis of Thermal Flux for Thermal Conduction of a Batch Chemical Reactor Equipped with a Monofluid Heating/Cooling System

    Directory of Open Access Journals (Sweden)

    Ghania Henini

    2012-01-01

    Full Text Available We present the thermal behavior of a batch reactor to jacket equipped with a monofluid heating/cooling system. Heating and cooling are provided respectively by an electrical resistance and two plate heat exchangers. The control of the temperature of the reaction is based on the supervision system. This strategy of management of the thermal devices is based on the usage of the thermal flux as manipulated variable. The modulation of the monofluid temperature by acting on the heating power or on the opening degrees of an air-to-open valve that delivers the monofluid to heat exchanger. The study shows that the application of this method for the conduct of the pilot reactor gives good results in simulation and that taking into account the dynamics of the various apparatuses greatly improves ride quality of conduct. In addition thermal control of an exothermic reaction (mononitration shows that the consideration of heat generated in the model representation improve the results by elimination any overshooting of the set-point temperature.

  11. CARDIOGRAMA: a stochastic, semi-empirical methodology for power-reactor surveillance and diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.A.; deSaussure, G.; Perez, R.B.

    1981-01-01

    The utilization of stochastic methods (reactor noise) for power reactor diagnostics and surveillance applications is by now a relatively well-established technique. In this technique, the power spectral density (PSD) of the fluctuations of a specified state variable is often used to define the reactor's signature at a given configuration. The purpose of the present work is to address the problem of handling efficiently the substantial amount of information involved in the application of reactor surveillance and diagnostics methods. Specifically, a methodology is described for: (a) representing the PSDs parametrically, and (b) detecting changes from the reactor's baseline PSD (normal signature).

  12. Some new viewpoints in reactor noise analysis

    Institute of Scientific and Technical Information of China (English)

    罗征培; 李富; 等

    1996-01-01

    It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model's phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.

  13. Risk analysis methodology survey

    Science.gov (United States)

    Batson, Robert G.

    1987-01-01

    NASA regulations require that formal risk analysis be performed on a program at each of several milestones as it moves toward full-scale development. Program risk analysis is discussed as a systems analysis approach, an iterative process (identification, assessment, management), and a collection of techniques. These techniques, which range from simple to complex network-based simulation were surveyed. A Program Risk Analysis Handbook was prepared in order to provide both analyst and manager with a guide for selection of the most appropriate technique.

  14. METHODOLOGICAL ELEMENTS OF SITUATIONAL ANALYSIS

    Directory of Open Access Journals (Sweden)

    Tetyana KOVALCHUK

    2016-07-01

    Full Text Available The article deals with the investigation of theoretical and methodological principles of situational analysis. The necessity of situational analysis is proved in modern conditions. The notion “situational analysis” is determined. We have concluded that situational analysis is a continuous system study which purpose is to identify dangerous situation signs, to evaluate comprehensively such signs influenced by a system of objective and subjective factors, to search for motivated targeted actions used to eliminate adverse effects of the exposure of the system to the situation now and in the future and to develop the managerial actions needed to bring the system back to norm. It is developed a methodological approach to the situational analysis, its goal is substantiated, proved the expediency of diagnostic, evaluative and searching functions in the process of situational analysis. The basic methodological elements of the situational analysis are grounded. The substantiation of the principal methodological elements of system analysis will enable the analyst to develop adaptive methods able to take into account the peculiar features of a unique object which is a situation that has emerged in a complex system, to diagnose such situation and subject it to system and in-depth analysis, to identify risks opportunities, to make timely management decisions as required by a particular period.

  15. Neutron detection of the Triga Mark III reactor, using nuclear track methodology

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

  16. Neutron detection of the Triga Mark III reactor, using nuclear track methodology

    Science.gov (United States)

    Espinosa, G.; Golzarri, J. I.; Raya-Arredondo, R.; Cruz-Galindo, S.; Sajo-Bohus, L.

    2015-07-01

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 1012 n cm-2 s-1, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer® PADC as neutron detection material, covered by 3 mm Plexiglas® as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

  17. Innovative fusion reactor design analysis: Annual performance report, May 15, 1988--January 31, 1989

    Energy Technology Data Exchange (ETDEWEB)

    Klein, A.C.

    1989-01-31

    This report discusses the following topics on fusion reactor component design: FLiBe intermediate heat exchanger design analysis; FLiBe properties; design methodology; FLiBe system steam generator freezeup; FLiBe reactor systems studies; tritium breeding ratio control; analysis of original objectives; and budget analysis. 15 refs., 13 figs., 3 tabs. (LSP)

  18. METHODOLOGICAL ELEMENTS OF SITUATIONAL ANALYSIS

    OpenAIRE

    Tetyana KOVALCHUK

    2016-01-01

    The article deals with the investigation of theoretical and methodological principles of situational analysis. The necessity of situational analysis is proved in modern conditions. The notion “situational analysis” is determined. We have concluded that situational analysis is a continuous system study which purpose is to identify dangerous situation signs, to evaluate comprehensively such signs influenced by a system of objective and subjective factors, to search for motivated targeted action...

  19. Application of Decomposition Methodology to Solve Integrated Process Design and Controller Design Problems for Reactor-Separator-Recycle System

    DEFF Research Database (Denmark)

    Abd.Hamid, Mohd-Kamaruddin; Sin, Gürkan; Gani, Rafiqul

    2010-01-01

    This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre-analysi...... to the solution of IPDC problems for RSR systems.......This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre...... the design of a RSR system involving consecutive reactions, A B -> C and shown to provide effective solutions that satisfy design, control and cost criteria. The advantage of the proposed methodology is that it is systematic, makes use of thermodynamic-process knowledge and provides valuable insights...

  20. PCCF flow analysis -- DR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Calkin, J.F.

    1961-04-26

    This report contains an analysis of PCCF tube flow and Panellit pressure relations at DR reactor. Supply curves are presented at front header pressures from 480 to 600 psig using cold water and the standard 0.236 inch orifice with taper down stream and the pigtail valve (plug or ball) open. Demand curves are presented for slug column lengths of 200 inches to 400 inches using 1.44 inch O.D. solid poison pieces (either Al or Pb-Cd) and cold water with a rear header pressure of 50 psig. Figure 1 is a graph of Panellit pressure vs. flow with the above supply and demand curves and clearly shows the effect of front header pressure and charge length on flow.

  1. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  2. Versatile in situ gas analysis apparatus for nanomaterials reactors.

    Science.gov (United States)

    Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

    2014-09-02

    We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

  3. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  4. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  5. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    Directory of Open Access Journals (Sweden)

    Giuseppe Palmiotti

    2012-01-01

    Full Text Available The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

  6. Thermal Analysis for Mobile Reactor

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>Mobile reactor design in the paper is consisted of two grades of thermal electric conversion. The first grade is the thermionic conversion inside the core and the second grade is thermocouple conversion

  7. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji [Malaysian Nuclear Agency, MOSTI, Bangi, 43000 Kajang, Selangor (Malaysia); Brayon, Fedrick Charlie Matthew [Atomic Energy Licensing Board, MOSTI, 43800 Dengkil, Selangor (Malaysia); Mohamed, Faizal [Universiti Kebangsaan Malaysia, Bangi, Selangor (Malaysia)

    2014-02-12

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  8. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    Science.gov (United States)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji

    2014-02-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  9. Mathematical Modeling for Simulation of Nuclear Reactor Analysis

    OpenAIRE

    Salah Ud-Din Khan; Shahab Ud-Din Khan

    2013-01-01

    In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...

  10. A methodology for modeling photocatalytic reactors for indoor pollution control using previously estimated kinetic parameters

    Energy Technology Data Exchange (ETDEWEB)

    Passalia, Claudio; Alfano, Orlando M. [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina); Brandi, Rodolfo J., E-mail: rbrandi@santafe-conicet.gov.ar [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer Indoor pollution control via photocatalytic reactors. Black-Right-Pointing-Pointer Scaling-up methodology based on previously determined mechanistic kinetics. Black-Right-Pointing-Pointer Radiation interchange model between catalytic walls using configuration factors. Black-Right-Pointing-Pointer Modeling and experimental validation of a complex geometry photocatalytic reactor. - Abstract: A methodology for modeling photocatalytic reactors for their application in indoor air pollution control is carried out. The methodology implies, firstly, the determination of intrinsic reaction kinetics for the removal of formaldehyde. This is achieved by means of a simple geometry, continuous reactor operating under kinetic control regime and steady state. The kinetic parameters were estimated from experimental data by means of a nonlinear optimization algorithm. The second step was the application of the obtained kinetic parameters to a very different photoreactor configuration. In this case, the reactor is a corrugated wall type using nanosize TiO{sub 2} as catalyst irradiated by UV lamps that provided a spatially uniform radiation field. The radiative transfer within the reactor was modeled through a superficial emission model for the lamps, the ray tracing method and the computation of view factors. The velocity and concentration fields were evaluated by means of a commercial CFD tool (Fluent 12) where the radiation model was introduced externally. The results of the model were compared experimentally in a corrugated wall, bench scale reactor constructed in the laboratory. The overall pollutant conversion showed good agreement between model predictions and experiments, with a root mean square error less than 4%.

  11. Analysis of Adiabatic Batch Reactor

    Directory of Open Access Journals (Sweden)

    Erald Gjonaj

    2016-05-01

    Full Text Available A mixture of acetic anhydride is reacted with excess water in an adiabatic batch reactor to form an exothermic reaction. The concentration of acetic anhydride and the temperature inside the adiabatic batch reactor are calculated with an initial temperature of 20°C, an initial temperature of 30°C, and with a cooling jacket maintaining the temperature at a constant of 20°C. The graphs of the three different scenarios show that the highest temperatures will cause the reaction to occur faster.

  12. Composite nuclear fuel fabrication methodology for gas fast reactors

    Science.gov (United States)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  13. New systematic methodology for incorporating dynamic heat transfer modelling in multi-phase biochemical reactors.

    Science.gov (United States)

    Fernández-Arévalo, T; Lizarralde, I; Grau, P; Ayesa, E

    2014-09-01

    This paper presents a new modelling methodology for dynamically predicting the heat produced or consumed in the transformations of any biological reactor using Hess's law. Starting from a complete description of model components stoichiometry and formation enthalpies, the proposed modelling methodology has integrated successfully the simultaneous calculation of both the conventional mass balances and the enthalpy change of reaction in an expandable multi-phase matrix structure, which facilitates a detailed prediction of the main heat fluxes in the biochemical reactors. The methodology has been implemented in a plant-wide modelling methodology in order to facilitate the dynamic description of mass and heat throughout the plant. After validation with literature data, as illustrative examples of the capability of the methodology, two case studies have been described. In the first one, a predenitrification-nitrification dynamic process has been analysed, with the aim of demonstrating the easy integration of the methodology in any system. In the second case study, the simulation of a thermal model for an ATAD has shown the potential of the proposed methodology for analysing the effect of ventilation and influent characterization.

  14. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  15. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  16. Design of a rotary reactor for chemical-looping combustion. Part 1: Fundamentals and design methodology

    KAUST Repository

    Zhao, Zhenlong

    2014-04-01

    Chemical-looping combustion (CLC) is a novel and promising option for several applications including carbon capture (CC), fuel reforming, H 2 generation, etc. Previous studies demonstrated the feasibility of performing CLC in a novel rotary design with micro-channel structures. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet, and depleted air and product streams at exit. The rotary wheel consists of a large number of micro-channels with oxygen carriers (OC) coated on the inner surface of the channel walls. In the CC application, the OC oxidizes the fuel while the channel is in the fuel zone to generate undiluted CO2, and is regenerated while the channel is in the air zone. In this two-part series, the effect of the reactor design parameters is evaluated and its performance with different OCs is compared. In Part 1, the design objectives and criteria are specified and the key parameters controlling the reactor performance are identified. The fundamental effects of the OC characteristics, the design parameters, and the operating conditions are studied. The design procedures are presented on the basis of the relative importance of each parameter, enabling a systematic methodology of selecting the design parameters and the operating conditions with different OCs. Part 2 presents the application of the methodology to the designs with the three commonly used OCs, i.e., nickel, copper, and iron, and compares the simulated performances of the designs. © 2013 Elsevier Ltd. All rights reserved.

  17. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    OpenAIRE

    Giuseppe Palmiotti; Massimo Salvatores

    2012-01-01

    The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the metho...

  18. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    OpenAIRE

    Masaki Kurata

    2016-01-01

    Research and development (R&D) methodology for the practical use of accident tolerant fuel (ATF) in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening proces...

  19. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

  20. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  1. Causal Meta-Analysis : Methodology and Applications

    NARCIS (Netherlands)

    Bax, L.J.

    2009-01-01

    Meta-analysis is a statistical method to summarize research data from multiple studies in a quantitative manner. This dissertation addresses a number of methodological topics in causal meta-analysis and reports the development and validation of meta-analysis software. In the first (methodological) p

  2. Reaction kinetic analysis of reactor surveillance data

    Science.gov (United States)

    Yoshiie, T.; Kinomura, A.; Nagai, Y.

    2017-02-01

    In the reactor pressure vessel surveillance data of a European-type pressurized water reactor (low-Cu steel), it was found that the concentration of matrix defects was very high, and a large number of precipitates existed. In this study, defect structure evolution obtained from surveillance data was simulated by reaction kinetic analysis using 15 rate equations. The saturation of precipitation and the growth of loops were simulated, but it was not possible to explain the increase in DBTT on the basis of the defect structures. The sub-grain boundary segregation of solutes was discussed for the origin of the DBTT increase.

  3. Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V. [JSC ' Atomtechenergo' , Filial ' Novovoronezhatomtechenergo' , Novovorenezh (Russian Federation); Ryasny, Sergei I. [JSC ' Atomtechenergo' , Mytishi (Russian Federation)

    2016-09-15

    At WWER-1000 NPPs, as well as at PWR NPPs, there is a problem of determining the correct weighted mean coolant temperature in the primary circuit hot legs based on the measuring channels information. The problem is caused by the coolant temperature stratification. The technical documentation for engineering support and maintenance of I and C systems does not provide any regulatory guidelines to consider this effect. Therefore, it is very important to represent a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of the WWER-1000 reactor plants. The given paper presents the basic preconditions and approaches applied during the methodology development. They were worked out on the basis of the executed numerical and experimental research taking into account the analysis of the extensive material obtained by the authors from full-scale tests during the commissioning of WWER-1000 power units, as well as operational data obtained from several power units with different fuel loadings.

  4. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part I, deterministic approaches

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang Kyu [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    The objective of this paper and a companion paper in this issue (part II, risk-informed approaches) is to derive technical insights from a critical review of deterministic and risk-informed safety analysis approaches that have been applied to develop licensing requirements for water-cooled reactors, or proposed for safety verification of the advanced reactor design. To this end, a review was made of a number of safety analysis approaches including those specified in regulatory guides and industry standards, as well as novel methodologies proposed for licensing of advanced reactors. This paper and the companion paper present the review insights on the deterministic and risk-informed safety analysis approaches, respectively. These insights could be used in making a safety case or developing a new licensing review infrastructure for advanced reactors including Generation IV reactors.

  5. Linear stability analysis of a nuclear reactor using the lumped model

    Directory of Open Access Journals (Sweden)

    Kale Vivek A.

    2016-01-01

    Full Text Available The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc. do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration.

  6. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  7. Methodology for Validating Building Energy Analysis Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Judkoff, R.; Wortman, D.; O' Doherty, B.; Burch, J.

    2008-04-01

    The objective of this report was to develop a validation methodology for building energy analysis simulations, collect high-quality, unambiguous empirical data for validation, and apply the validation methodology to the DOE-2.1, BLAST-2MRT, BLAST-3.0, DEROB-3, DEROB-4, and SUNCAT 2.4 computer programs. This report covers background information, literature survey, validation methodology, comparative studies, analytical verification, empirical validation, comparative evaluation of codes, and conclusions.

  8. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.

  9. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    Science.gov (United States)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  10. METHODOLOGICAL STRATEGIES FOR TEXTUAL DATA ANALYSIS:

    OpenAIRE

    Juan Carlos Rincón-Vásquez; Andrea Velandia-Morales; Idaly Barreto

    2011-01-01

    This paper presents a classification methodology for studies of textual data. Thisclassification is based on the two predominant methodologies for social scienceresearch: qualitative and quantitative. The basic assumption is that the researchprocess involves three main features: 1) Structure Research, 2) Collection of informationand, 3) Analysis and Interpretation of Data. In each, there are generalguidelines for textual studies.

  11. METHODOLOGICAL STRATEGIES FOR TEXTUAL DATA ANALYSIS:

    Directory of Open Access Journals (Sweden)

    Juan Carlos Rincón-Vásquez

    2011-12-01

    Full Text Available This paper presents a classification methodology for studies of textual data. Thisclassification is based on the two predominant methodologies for social scienceresearch: qualitative and quantitative. The basic assumption is that the researchprocess involves three main features: 1 Structure Research, 2 Collection of informationand, 3 Analysis and Interpretation of Data. In each, there are generalguidelines for textual studies.

  12. Constructive Analysis : A Study in Epistemological Methodology

    DEFF Research Database (Denmark)

    Ahlström, Kristoffer

    The present study is concerned the viability of the primary method in contemporary philosophy, i.e., conceptual analysis. Starting out by tracing the roots of this methodology to Platonic philosophy, the study questions whether such a methodology makes sense when divorced from Platonic philosophy...

  13. Reactivity determination in accelerator driven reactors using reactor noise analysis

    Directory of Open Access Journals (Sweden)

    Kostić Ljiljana 1

    2002-01-01

    Full Text Available Feynman-alpha and Rossi-alpha methods are used in traditional nuclear reactors to determine the subcritical reactivity of a system. The methods are based on the measurement of the mean value, variance and the covariance of detector counts for different measurement times. Such methods attracted renewed attention recently with the advent of the so-called accelerator driven reactors (ADS proposed some time ago. The ADS systems, intended to be used either in energy generation or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those traditionally used by radioactive sources. In such reactors the monitoring of the subcritical reactivity is very important, and a statistical method, such as the Feynman-alpha method, is capable of resolving this problem.

  14. Stress analysis of magnetically controlled reactor

    Directory of Open Access Journals (Sweden)

    Ben Tong

    2017-05-01

    Full Text Available To provide technique references for vibration reduction of magnetically controlled reactors (MCRs, stress, which is the inherent reason of vibration and noise, should be investigated. Stresses in reactor cores are produced due to the magnetostriction deformation of silicon steel and electromagnetic force between the core discs. So far, stress analysis on reactor cores was based on one-way coupled numerical method, which did not consider the influence of the stress on magnetic properties of the core material. Thus, multi-group magnetization and magnetostriction characteristics curves of silicon steel under different tensile stresses are measured firstly to support the computation. From the experiment results, it can be seen that magnetic properties of silicon steel change with stress. Then an electromagneto-mechanical two-way coupled numerical model for MCRs considering magnetostrictive effect and electromagnetic force effect is proposed. Stress distribution of MCR cores under the combination excitation of the sinusoidal wave current and different direct currents are calculated. From the computed results, it can be seen that a larger direct current has greater influence on MCRs vibration, which provides a theory basis for further analysis of vibration and noise reduction.

  15. Stress analysis of magnetically controlled reactor

    Science.gov (United States)

    Tong, Ben; Qingxin, Yang; Rongge, Yan; Lihua, Zhu; Ling, Weng; Ying, Sun

    2017-05-01

    To provide technique references for vibration reduction of magnetically controlled reactors (MCRs), stress, which is the inherent reason of vibration and noise, should be investigated. Stresses in reactor cores are produced due to the magnetostriction deformation of silicon steel and electromagnetic force between the core discs. So far, stress analysis on reactor cores was based on one-way coupled numerical method, which did not consider the influence of the stress on magnetic properties of the core material. Thus, multi-group magnetization and magnetostriction characteristics curves of silicon steel under different tensile stresses are measured firstly to support the computation. From the experiment results, it can be seen that magnetic properties of silicon steel change with stress. Then an electromagneto-mechanical two-way coupled numerical model for MCRs considering magnetostrictive effect and electromagnetic force effect is proposed. Stress distribution of MCR cores under the combination excitation of the sinusoidal wave current and different direct currents are calculated. From the computed results, it can be seen that a larger direct current has greater influence on MCRs vibration, which provides a theory basis for further analysis of vibration and noise reduction.

  16. Common-Cause Failure Analysis for Reactor Protection System Reliability Studies

    Energy Technology Data Exchange (ETDEWEB)

    Gentillon, C.; Rasmuson, D.; Eide, S.; Wierman, T.

    1999-08-01

    Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.

  17. Rat sperm motility analysis: methodologic considerations

    Science.gov (United States)

    The objective of these studies was to optimize conditions for computer-assisted sperm analysis (CASA) of rat epididymal spermatozoa. Methodologic issues addressed include sample collection technique, sampling region within the epididymis, type of diluent medium used, and sample c...

  18. Development of a simplified methodology for the isotopic determination of fuel spent in Light Water Reactors; Desarrollo de una metodologia simplificada para la determinacion isotopica del combustible gastado en reactores de agua ligera

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez N, H.; Francois L, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: hermilo@lairn.fi-b.unam.mx

    2005-07-01

    The present work presents a simplified methodology to quantify the isotopic content of the spent fuel of light water reactors; their application is it specific to the Laguna Verde Nucleo electric Central by means of a balance cycle of 18 months. The methodology is divided in two parts: the first one consists on the development of a model of a simplified cell, for the isotopic quantification of the irradiated fuel. With this model the burnt one is simulated 48,000 MWD/TU of the fuel in the core of the reactor, taking like base one fuel assemble type 10x10 and using a two-dimensional simulator for a fuel cell of a light water reactor (CPM-3). The second part of the methodology is based on the creation from an isotopic decay model through an algorithm in C++ (decay) to evaluate the amount, by decay of the radionuclides, after having been irradiated the fuel until the time in which the reprocessing is made. Finally the method used for the quantification of the kilograms of uranium and obtained plutonium of a normalized quantity (1000 kg) of fuel irradiated in a reactor is presented. These results will allow later on to make analysis of the final disposition of the irradiated fuel. (Author)

  19. Comparative analysis of seismic risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyoo; Kim, Tae Woon; Hwang, Mi Jung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    SRA methodologies are separated into SPSA and SMM. SPSA methodology that has been widely used for seismic risk analysis has two kinds of methodologies such as Zion method and SSMRP method. SPSA methodology is suitable to interfacing with the analysis of internal event. However, the results of SPSA have uncertainties because of uncertainties in seismic hazard analysis and subjective judgement. Zion method specially developed for commercial use is less expensive and less time consuming but more uncertain than SSMRP method, since the former performs the fragility analysis less in detail than the latter. SMM is impossible to interface with the analysis of internal event but the uncertainties that are occurred during seismic hazard analysis is reduced because of the screening using RLE (review level earthquake). Therefore, if SPSA-based SMM methodology is chosen to be developed, the results of SRA will be more reliable and it requires low costs and time. In addition, the new methodology will require the development of a new evaluating code for SRA. (Author) 26 refs., 25 figs., 16 tabs.

  20. Similar methodological analysis involving the user experience.

    Science.gov (United States)

    Almeida e Silva, Caio Márcio; Okimoto, Maria Lúcia R L; Tanure, Raffaela Leane Zenni

    2012-01-01

    This article deals with the use of a protocol for analysis of similar methodological analysis related to user experience. For both, were selected articles recounting experiments in the area. They were analyze based on the similar analysis protocol and finally, synthesized and associated.

  1. Steady State Analysis of Small Molten Salt Reactor : Effect of Fuel Salt Flow on Reactor Characteristics

    OpenAIRE

    Yamamoto, Takahisa; MITACHI, Koshi; Suzuki, Takashi

    2005-01-01

    The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fue...

  2. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  3. Human Reliability Analysis for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  4. Malware Analysis Sandbox Testing Methodology

    Directory of Open Access Journals (Sweden)

    Zoltan Balazs

    2016-01-01

    Full Text Available Manual processing of malware samples became impossible years ago. Sandboxes are used to automate the analysis of malware samples to gather information about the dynamic behaviour of the malware, both at AV companies and at enterprises. Some malware samples use known techniques to detect when it runs in a sandbox, but most of these sandbox detection techniques can be easily detected and thus flagged as malicious. I invented new approaches to detect these sandboxes. I developed a tool, which can collect a lot of interesting information from these sandboxes to create statistics how the current technologies work. After analysing these results I will demonstrate tricks to detect sandboxes. These tricks can’t be easily flagged as malicious. Some sandboxes don’t not interact with the Internet in order to block data extraction, but with some DNS-fu the information can be extracted from these appliances as well.

  5. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  6. Accidental safety analysis methodology development in decommission of the nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

    2002-03-15

    Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

  7. Integrated systems analysis of the PIUS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  8. Methodology comparison for gamma-heating calculations in material-testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)

    2015-07-01

    The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear

  9. Exploring participatory methodologies in organizational discourse analysis

    DEFF Research Database (Denmark)

    Plotnikof, Mie

    2014-01-01

    Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new...... in collaborative governance processes in Denmark. The paper contributes to refining methodologies of organizational discourse analysis by elaborating method-mixing that embraces multimodal organizational discourses. Furthermore it discusses practical implications of the struggling subjectification processes......-and practices by dealing with challenges of methodological overview, responsive creativity and identity-struggle. The potentials hereof are demonstrated and discussed with cases of two both critical and co-creative practices, namely ‘organizational modelling’ and ‘fixed/unfixed positioning’ from fieldwork...

  10. Exploring participatory methodologies in organizational discourse analysis

    DEFF Research Database (Denmark)

    Plotnikof, Mie

    2014-01-01

    Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new contribu......Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new...... contributions. As regards conceptual efforts are made but further exploration of methodological combinations and their practical implications are called for. This paper argues 1) to combine methodologies by approaching this as scholarly subjectification processes, and 2) to perform combinations in both......-and practices by dealing with challenges of methodological overview, responsive creativity and identity-struggle. The potentials hereof are demonstrated and discussed with cases of two both critical and co-creative practices, namely ‘organizational modelling’ and ‘fixed/unfixed positioning’ from fieldwork...

  11. Calculation methodology of the thermal margin in the CAREM 25 reactor; Metodologia de calculo del margen termico en el reactor CAREM 25

    Energy Technology Data Exchange (ETDEWEB)

    Mazufri, Claudio M. [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)

    1995-12-31

    According to the nuclear reactors characteristics, can be found different methodologies to appraise the thermal margin available in the core. In the particular case of the CAREM (25 MWe) reactor, where the core is cooled by low mass flux and there are zones with positive steam quality, such evaluation is critical. Due to these characteristics, it was necessary to develop one proper methodology. In the present work, the different steps of that development are described: the election of figures of merit for measure the thermal margin, the hypothesis to use, the election of the critical heat flux prediction model, model qualification and the specification of the core wide procedure. In each step assume criteria are discussed. (author). 9 refs, 1 tab, 1 fig.

  12. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

  13. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  14. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  15. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  16. Fractal analysis: methodologies for biomedical researchers.

    Science.gov (United States)

    Ristanović, Dusan; Milosević, Nebojsa T

    2012-01-01

    Fractal analysis has become a popular method in all branches of scientific investigations including biology and medicine. Although there is a growing interest in the application of fractal analysis in biological sciences, questions about the methodology of fractal analysis have partly restricted its wider and comprehensible application. It is a notable fact that fractal analysis is derived from fractal geometry, but there are some unresolved issues that need to be addressed. In this respect, we discuss several related underlying principles for fractal analysis and establish the meaningful relationship between fractal analysis and fractal geometry. Since some concepts in fractal analysis are determined descriptively and/or qualitatively, this paper provides their exact mathematical definitions or explanations. Another aim of this study is to show that nowadays fractal analysis is an independent mathematical and experimental method based on Mandelbrot's fractal geometry, Euclidean traditiontal geometry and Richardson's coastline method.

  17. Using of BEPU methodology in a final safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Menzel, Francine; Sabundjian, Gaiane, E-mail: fmenzel@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); D' auria, Francesco, E-mail: f.dauria@ing.unipi.it [Universita degli Studi di Pisa, Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG), Pisa (Italy); Madeira, Alzira A., E-mail: alzira@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

  18. Ion beam analysis of materials in the PBMR reactor

    Science.gov (United States)

    Malherbe, Johan B.; Friedland, E.; van der Berg, N. G.

    2008-04-01

    South Africa is developing a new type of high temperature nuclear reactor, the so-called pebble bed modular reactor (PBMR). The planned reactor outlet temperature of this gas-cooled reactor is approximately 900 °C. This high temperature places some severe restrictions on materials, which can be used. The name of the reactor is derived from the form of the fuel elements, which are in the form of pebbles, each with a diameter of 60 mm. Each pebble is composed of several thousands of coated fuel particles. The coated particle consists of a nucleus of UO2 surrounded by several layers of different carbons and SiC. The diameter of the fuel particles is 0.92 mm. A brief review will be given of the advantages of this nuclear reactor, of the materials in the fuel elements and their analysis using ion beam techniques.

  19. An Analysis of Software Design Methodologies

    Science.gov (United States)

    1979-08-01

    Technical Report 401-, # "AN ANALYSIS OF SOFTWARE DESIGN METHODOLOGIES H. Rudy Ramsey, Michael E. Atwood , and Gary D. Campbell Science...H. Rudy Ramsey, Michael E. Atwood , and Gary D. Campbell Science Applications, Incorporated Submitted by: Edgar M. Johnson, Chief HUMAN FACTORS...expressed by members ot the Integrated Software Research and Development Working Group (ISRAD). The authors are indebted to Martha Cichelli, Margaret

  20. Development of fault diagnostic technique using reactor noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  1. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    Directory of Open Access Journals (Sweden)

    Masaki Kurata

    2016-02-01

    Full Text Available Research and development (R&D methodology for the practical use of accident tolerant fuel (ATF in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening process among stakeholders in the near term and in developing an efficient R&D track toward practical use. Technology readiness levels and attribute guides are considered to be proper indices for these evaluations. In the midterm, the selected ATF-concepts will be developed toward the technology readiness level-5, at which stage the performance of the prototype fuel rods and the practicality of industrial scale fuel manufacturing will be verified and validated. Regarding the screened-out concepts, which are recognized to have attractive potentials, the fundamental R&D should be continued in the midterm to find ways of addressing showstoppers.

  2. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  3. Assessing Pretreatment Reactor Scaling Through Empirical Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; Nagle, Nicholas J.; Schell, Daniel J.; Tucker, Melvin P.; McMillan, James D.; Wolfrum, Edward J.

    2016-12-01

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, this is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was

  4. Mass Spectrometry Methodology in Lipid Analysis

    Directory of Open Access Journals (Sweden)

    Lin Li

    2014-06-01

    Full Text Available Lipidomics is an emerging field, where the structures, functions and dynamic changes of lipids in cells, tissues or body fluids are investigated. Due to the vital roles of lipids in human physiological and pathological processes, lipidomics is attracting more and more attentions. However, because of the diversity and complexity of lipids, lipid analysis is still full of challenges. The recent development of methods for lipid extraction and analysis and the combination with bioinformatics technology greatly push forward the study of lipidomics. Among them, mass spectrometry (MS is the most important technology for lipid analysis. In this review, the methodology based on MS for lipid analysis was introduced. It is believed that along with the rapid development of MS and its further applications to lipid analysis, more functional lipids will be identified as biomarkers and therapeutic targets and for the study of the mechanisms of disease.

  5. Hygro-Thermo-Mechanical Analysis of a Reactor Vessel

    Directory of Open Access Journals (Sweden)

    Jaroslav Kruis

    2012-01-01

    Full Text Available Determining the durability of a reactor vessel requires a hygro-thermo-mechanical analysis of the vessel throughout its service life. Damage, prestress losses, distribution of heat and moisture and some other quantities are needed for a durability assessment. A coupled analysis was performed on a two-level model because of the huge demands on computer hardware. This paper deals with a hygro-thermo-mechanical analysis of a reactor vessel made of prestressed concrete with a steel inner liner. The reactor vessel is located in Temelín, Czech Republic.

  6. Multiscale Methods for Nuclear Reactor Analysis

    Science.gov (United States)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  7. Theoretical and methodological approaches in discourse analysis.

    Science.gov (United States)

    Stevenson, Chris

    2004-10-01

    Discourse analysis (DA) embodies two main approaches: Foucauldian DA and radical social constructionist DA. Both are underpinned by social constructionism to a lesser or greater extent. Social constructionism has contested areas in relation to power, embodiment, and materialism, although Foucauldian DA does focus on the issue of power. Embodiment and materialism may be especially relevant for researchers of nursing where the physical body is prominent. However, the contested nature of social constructionism allows a fusion of theoretical and methodological approaches tailored to a specific research interest. In this paper, Chris Stevenson suggests a frame- work for working out and declaring the DA approach to be taken in relation to a research area, as well as to aid anticipating methodological critique. Method, validity, reliability and scholarship are discussed from within a discourse analytic frame of reference.

  8. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  9. Application of Decomposition Methodology to Solve Integrated Process Design and Controller Design Problems for Reactor-Separator-Recycle System

    DEFF Research Database (Denmark)

    Abd.Hamid, Mohd-Kamaruddin; Sin, Gürkan; Gani, Rafiqul

    2010-01-01

    This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre...... the design of a RSR system involving consecutive reactions, A B -> C and shown to provide effective solutions that satisfy design, control and cost criteria. The advantage of the proposed methodology is that it is systematic, makes use of thermodynamic-process knowledge and provides valuable insights...

  10. Multiscale Analysis of Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar; Woo Yoon; Abderrafi Ougouag

    2010-10-01

    – The PEBBED code was developed at the Idaho National Laboratory for design and analysis of pebble-bed high temperature reactors. The diffusion-depletion-pebble-mixing algorithm of the original PEBBED code was enhanced through coupling with the THERMIX-KONVEK code for thermal fluid analysis and by the COMBINE code for online cross section generation. The COMBINE code solves the B-1 or B-3 approximations to the transport equation for neutron slowing down and resonance interactions in a homogeneous medium with simple corrections for shadowing and thermal self-shielding. The number densities of materials within specified regions of the core are averaged and transferred to COMBINE from PEBBED for updating during the burnup iteration. The simple treatment of self-shielding in previous versions of COMBINE led to inaccurate results for cross sections and unsatisfactory core performance calculations. A new version of COMBINE has been developed that treats all levels of heterogeneity using the 1D transport code ANISN. In a 3-stage calculation, slowing down is performed in 167 groups for each homogeneous subregion (kernel, particle layers, graphite shell, control rod absorber annulus, etc.) Particles in a local average pebble are homogenized using ANISN then passed to the next (pebble) stage. A 1D transport solution is again performed over the pebble geometry and the homogenized pebble cross sections are passed to a 1-d radial model of a wedge of the pebble bed core. This wedge may also include homogeneous reflector regions and a control rod region composed of annuli of different absorbing regions. Radial leakage effects are therefore captured with discrete ordinates transport while axial and azimuthal effects are captured with a transverse buckling term. In this paper, results of various PBR models will be compared with comparable models from literature. Performance of the code will be assessed.

  11. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Downar; E. Lewis

    2005-08-31

    Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

  12. RAMS (Risk Analysis - Modular System) methodology

    Energy Technology Data Exchange (ETDEWEB)

    Stenner, R.D.; Strenge, D.L.; Buck, J.W. [and others

    1996-10-01

    The Risk Analysis - Modular System (RAMS) was developed to serve as a broad scope risk analysis tool for the Risk Assessment of the Hanford Mission (RAHM) studies. The RAHM element provides risk analysis support for Hanford Strategic Analysis and Mission Planning activities. The RAHM also provides risk analysis support for the Hanford 10-Year Plan development activities. The RAMS tool draws from a collection of specifically designed databases and modular risk analysis methodologies and models. RAMS is a flexible modular system that can be focused on targeted risk analysis needs. It is specifically designed to address risks associated with overall strategy, technical alternative, and `what if` questions regarding the Hanford cleanup mission. RAMS is set up to address both near-term and long-term risk issues. Consistency is very important for any comparative risk analysis, and RAMS is designed to efficiently and consistently compare risks and produce risk reduction estimates. There is a wide range of output information that can be generated by RAMS. These outputs can be detailed by individual contaminants, waste forms, transport pathways, exposure scenarios, individuals, populations, etc. However, they can also be in rolled-up form to support high-level strategy decisions.

  13. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

    2016-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.

  14. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor

    2017-03-01

    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  15. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

    2017-03-15

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  16. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  17. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  18. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  19. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  20. NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21

    Energy Technology Data Exchange (ETDEWEB)

    Smith, G L

    1989-12-01

    The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.

  1. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

    Directory of Open Access Journals (Sweden)

    A. Rais

    2015-01-01

    Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

  2. Seismic Base Isolation Analysis for PASCAR Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kuk Hee; Yoo, Bong; Kim, Yun Jae [Korea Univ., Seoul (Korea, Republic of)

    2008-10-15

    This paper presents a study for developing a seismic isolation system for the PASCAR (Proliferation resistant, Accident-tolerant, Self-supported, Capsular and Assured Reactor) liquid metal reactor design. PASCAR use lead-bismuth eutectic (LBE) as coolant. Because the density (10,000kg/m{sup 3}) of LBE coolant is very heavier than sodium coolant and water, this presents a challenge to designers of the seismic isolation systems that will be used with these heavy liquid metal reactors. Finite element analysis is adapted to determine the characteristics of the isolator device. Results are presented from a study on the use of three-dimensional seismic isolation devices to the full-scale reactor. The seismic analysis responses of the two-dimensional and the three-dimensional isolation systems for the PASCAR are compared with that of the conventional fixed base system.

  3. Parameter analysis calculation on characteristics of portable FAST reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  4. Methodology of on-line prediction and fuel management of the modular pebble-bed high-temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bing, X.; Fu, L.; Yanhua, Z.; Fubing, C.; Jian, Z.; Jiong, G., E-mail: xiabing@tsinghua.edu.cn [Tsinghua Univ., Inst. of Nuclear and New Energy Technology (INET), Beijing (China)

    2014-07-01

    The modular pebble-bed high-temperature reactors (PB-HTRs), featured by the inherent safety and modular design concept, are considered as one of the promising candidates for the nuclear system of next generation. The characteristics of on-line successive fueling, as well as the features of small excess reactivity and lack of reactivity control methods, make the fuel management and operation of PB-HTRs coupled tightly with each other. The on-line fuel management of PB-HTRs needs the capability of on-line prediction for future operation scenarios based on the combination of several tightly coupled key parameters of the reactor core, including the power level of reactor core, the unloading speed and loading ratio of fuel/dummy pebbles, and the control rod positions. The methodology of on-line prediction is proposed, and then verified by the model of the HTR-10, a small test modular PB-HTR with nominal power of 10 MW. For different combinations, the prediction sequences are calculated by using the computer code system of VSOP for both equilibrium state and running-in phase, especially the latter. The prediction results are analyzed by using a series of data processing based on the polynomial interpolation to determine the optimized parameters for fuel management and core operation of next step. The verification of methodology on the HTR-10 model demonstrates the feasibility of the on-line prediction and fuel management of modular PH-HTRs. (author)

  5. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  6. Impact of inflow transport approximation on light water reactor analysis

    Science.gov (United States)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  7. Cost analysis methodology: Photovoltaic Manufacturing Technology Project

    Energy Technology Data Exchange (ETDEWEB)

    Whisnant, R.A. (Research Triangle Inst., Research Triangle Park, NC (United States))

    1992-09-01

    This report describes work done under Phase 1 of the Photovoltaic Manufacturing Technology (PVMaT) Project. PVMaT is a five-year project to support the translation of research and development in PV technology into the marketplace. PVMaT, conceived as a DOE/industry partnership, seeks to advanced PV manufacturing technologies, reduce PV module production costs, increase module performance, and expand US commercial production capacities. Under PVMaT, manufacturers will propose specific manufacturing process improvements that may contribute to the goals of the project, which is to lessen the cost, thus hastening entry into the larger scale, grid-connected applications. Phase 1 of the PVMaT project is to identify obstacles and problems associated with manufacturing processes. This report describes the cost analysis methodology required under Phase 1 that will allow subcontractors to be ranked and evaluated during Phase 2.

  8. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  9. Analysis of Credible Accidents for Argonaut Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hawley, S. C.; Kathern, R. L.; Robkin, M. A.

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.

  10. Application of response surface methodology to optimize the operational parameters for enhanced removal efficiency of organic matter and nitrogen: moving bed biofilm reactor.

    Science.gov (United States)

    Barwal, Anjali; Chaudhary, Rubina

    2016-05-01

    An attempt of response surface methodology (RSM) has been made for more effective utilization and optimization for considerable reduction of operational conditions such as reaction time, aeration time, energy consumption, etc. for municipal wastewater treatment process using moving bed biofilm reactor (MBBR). A mathematical-statistical model was developed for the second-order response surface through the fit of a polynomial function and a central composite design (CCD) in the form of a full factorial design. CCD was employed to assess the interactive effects of the three main independent operational parameters, including biocarrier filling rate (0-70 %), aeration rate (0.21-0.42 m(3) h(-1)), and reactor run time (1-15 days), on the removal efficiency of chemical oxygen demand (COD), biochemical oxygen demand (BOD), and total Kjeldahl nitrogen (TKN). Analysis of variance expressed a high coefficient of determination (R (2) = 0.84-0.95), thereby indicating that the model is significant. Using a desirability function for the highest COD (93 %), BOD (96 %), and TKN (69 %) removal, the optimum carrier filling rate, aeration rate, and reactor run time were identified to be 40 %, 0.21 m(3) h(-1), and 7 days, respectively. It shows that RSM can be a suitable method to optimize the operational parameters of MBBR with enhanced removal efficiency and less power consumption.

  11. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  12. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  13. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  14. Production of specifically structured lipids by enzymatic interesterification in a pilot enzyme bed reactor: process optimization by response surface methodology

    DEFF Research Database (Denmark)

    Xu, Xuebing; Mu, Huiling; Høy, Carl-Erik;

    1999-01-01

    Pilot production of specifically structured lipids by Lipozyme IM-catalyzed interesterification was carried out in a continuous enzyme bed reactor without the use of solvent. Medium chain triacylglycerols and oleic acid were used as model substrates. Response surface methodology was applied...... and the production of mono-incorporated and di-incorporated structured lipids with multiple regression and backward elimination. The coefficient of determination (R2) for the incorporation was 0.93, and that for the di-incorporated products was 0.94. The optimal conditions were flow rate, 2 ml/min; temperature, 65...

  15. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  16. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  17. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems: Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well establish

  18. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems: Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

  19. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  20. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  1. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  2. Nonlinear analysis and prediction of time series in multiphase reactors

    CERN Document Server

    Liu, Mingyan

    2014-01-01

    This book reports on important nonlinear aspects or deterministic chaos issues in the systems of multi-phase reactors. The reactors treated in the book include gas-liquid bubble columns, gas-liquid-solid fluidized beds and gas-liquid-solid magnetized fluidized beds. The authors take pressure fluctuations in the bubble columns  as time series for nonlinear analysis, modeling and forecasting. They present qualitative and quantitative non-linear analysis tools which include attractor phase plane plot, correlation dimension, Kolmogorov entropy and largest Lyapunov exponent calculations and local non-linear short-term prediction.

  3. Clean Energy Manufacturing Analysis Center Benchmark Report: Framework and Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Sandor, Debra [National Renewable Energy Lab. (NREL), Golden, CO (United States); Chung, Donald [National Renewable Energy Lab. (NREL), Golden, CO (United States); Keyser, David [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mann, Margaret [National Renewable Energy Lab. (NREL), Golden, CO (United States); Engel-Cox, Jill [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-05-23

    This report documents the CEMAC methodologies for developing and reporting annual global clean energy manufacturing benchmarks. The report reviews previously published manufacturing benchmark reports and foundational data, establishes a framework for benchmarking clean energy technologies, describes the CEMAC benchmark analysis methodologies, and describes the application of the methodologies to the manufacturing of four specific clean energy technologies.

  4. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  5. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  6. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  7. Application of cluster analysis and autoregressive neural networks for the noise diagnostics of the IBR-2M reactor

    Science.gov (United States)

    Pepelyshev, Yu. N.; Tsogtsaikhan, Ts.; Ososkov, G. A.

    2016-09-01

    The pattern recognition methodologies and artificial neural networks were used widely for the IBR-2M pulsed reactor noise diagnostics. The cluster analysis allows a detailed study of the structure and fast reactivity effects of IBR-2M and nonlinear autoregressive neural network (NAR) with local feedback connection allows predicting slow reactivity effects. In this work we present results of a study on pulse energy noise dynamics and prediction of liquid sodium flow rate through the core of the IBR-2M reactor using cluster analysis and an artificial neural network.

  8. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  9. Utility guidelines for reactor noise analysis: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sweeney, F.J.

    1987-02-01

    Noise analysis techniques have been extensively utilized to monitor the health and performance of nuclear power plant systems. However, few utilities have adequate programs to effectively utilize these techniques. These programs usually provide low-quality data, which can lead to misinterpretation and false alarms. The objective of this work is to provide utilities and noise analysts with guidelines for data acquisition, data analysis, and interpretation of noise analysis results for surveillance and diagnosis of reactor systems.

  10. Methodology to evaluate the crack growth rate by stress corrosion cracking in dissimilar metals weld in simulated environment of PWR nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Paula, Raphael G.; Figueiredo, Celia A.; Rabelo, Emerson G., E-mail: raphaelmecanica@gmail.com, E-mail: caf@cdtn.br, E-mail: egr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    Inconel alloys weld metal is widely used to join dissimilar metals in nuclear reactors applications. It was recently observed failures of weld components in plants, which have triggered an international effort to determine reliable data on the stress corrosion cracking behavior of this material in reactor environment. The objective of this work is to develop a methodology to determine the crack growth rate caused by stress corrosion in Inconel alloy 182, using the specimen (Compact Tensile) in simulated PWR environment. (author)

  11. Deposition reactors for solar grade silicon: A comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

    Science.gov (United States)

    Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.

    2015-12-01

    Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.

  12. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  13. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  14. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  15. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  16. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  17. The methodology of semantic analysis for extracting physical effects

    Science.gov (United States)

    Fomenkova, M. A.; Kamaev, V. A.; Korobkin, D. M.; Fomenkov, S. A.

    2017-01-01

    The paper represents new methodology of semantic analysis for physical effects extracting. This methodology is based on the Tuzov ontology that formally describes the Russian language. In this paper, semantic patterns were described to extract structural physical information in the form of physical effects. A new algorithm of text analysis was described.

  18. Using a Realist Research Methodology in Policy Analysis

    Science.gov (United States)

    Lourie, Megan; Rata, Elizabeth

    2017-01-01

    The article describes the usefulness of a realist methodology in linking sociological theory to empirically obtained data through the development of a methodological device. Three layers of analysis were integrated: 1. the findings from a case study about Maori language education in New Zealand; 2. the identification and analysis of contradictions…

  19. Using a Realist Research Methodology in Policy Analysis

    Science.gov (United States)

    Lourie, Megan; Rata, Elizabeth

    2017-01-01

    The article describes the usefulness of a realist methodology in linking sociological theory to empirically obtained data through the development of a methodological device. Three layers of analysis were integrated: 1. the findings from a case study about Maori language education in New Zealand; 2. the identification and analysis of contradictions…

  20. 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.

  1. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  2. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  3. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  4. Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Williams, Brian [Los Alamos National Laboratory; Mc Clure, Patrick [Los Alamos National Laboratory; Nelson, Ralph A [IDAHO NATIONAL LAB

    2010-01-01

    Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M&S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for cost

  5. Production of specifically structured lipids by enzymatic interesterification in a pilot enzyme bed reactor: process optimization by response surface methodology

    DEFF Research Database (Denmark)

    Xu, Xuebing; Mu, Huiling; Høy, Carl-Erik

    1999-01-01

    to optimize the reaction system with four process parameters, these being volume flow rate, water content in the substrates, reaction temperature and substrate ratio. The incorporation of acyl donors, product yields and the content of diacylglycerols were measured as the model responses. Enzyme activity......Pilot production of specifically structured lipids by Lipozyme IM-catalyzed interesterification was carried out in a continuous enzyme bed reactor without the use of solvent. Medium chain triacylglycerols and oleic acid were used as model substrates. Response surface methodology was applied...... was not identical for the sequential experiments in the same enzyme bed due to the deactivation of the Lipozyme IM. Therefore, the results were normalized based on enzyme deactivation models. Well-fitting quadratic models were obtained after normalizing the data for the incorporation of oleic acid...

  6. Methodology for risk analysis of projects

    Directory of Open Access Journals (Sweden)

    Roșu Maria Magdalena

    2017-01-01

    Full Text Available The risk in the organization activity as an economic and social system open, adaptive, with varying degrees of permeability to the influences from the current business environment, which is increasingly unpredictable and in which the only constant is the change, refers to the probability of not complying with the objectives set in terms of performance, execution and cost. The insufficient application of the recognized project management methodologies can be one of the main causes of projects failures in the organization with major influences on the activity efficiency and the performance recorded. Therefore, the methodology proposed in the paper, wants to be an effective tool, a formalized risk management tool, considered as a cyclical process, with several distinct phases, indispensable to the current organizational practice which should contribute to optimizing the project performance and its successful completion.

  7. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  8. Accident progression event tree analysis for postulated severe accidents at N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

  9. Methodological Aspects Regarding The Organizational Stress Analysis

    Science.gov (United States)

    Irimie, Sabina; Pricope (Muntean), Luminiţa Doina; Pricope, Sorin; Irimie, Sabin Ioan

    2015-07-01

    This work presents a research of methodology in occupational stress analyse in the educational field, as a part of a larger study. The objectives of the work are in finding accents in existence of significant relations between stressors and effects, meaning the differences between the indicators of occupational stress to teaching staff in primary and gymnasium school, taking notice of each specific condition: the institution as an entity, the working community, the discipline he/she is teaching others, the geographic and administrative district (urban/rural) and the quantification of stress level.

  10. Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oztunali, O.I.; Roles, G.W.

    1986-01-01

    Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology.

  11. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    NARCIS (Netherlands)

    Sjenitzer, B.L.

    2013-01-01

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

  12. Qualitative diagnosis for transients analysis on nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lorre, J.P.; Dorlet, E.; Evrard, J.M.

    1995-12-31

    One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs.

  13. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized- water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SAS2H analytical sequence in SCALE-4. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE-4 criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for analysis of each critical configuration. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power

  14. A bibliography on finite element and related methods analysis in reactor physics computations (1971--1997)

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.C.

    1998-01-01

    This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.

  15. Dynamic modeling of primary and secondary systems of IRIS reactor for transient analysis using SIMULINK

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Lima, Fernando Roberto de Andrade, E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2011-07-01

    The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)

  16. Example Work Domain Analysis for a Reference Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.

  17. METHODOLOGICAL ANALYSIS OF TRAINING STUDENT basketball teams

    Directory of Open Access Journals (Sweden)

    Kozina Zh.L.

    2011-06-01

    Full Text Available Considered the leading position of the preparation of basketball teams in high schools. The system includes the following: reliance on top-quality players in the structure of preparedness, widespread use of visual aids, teaching movies and cartoons with a record of technology implementation of various methods by professional basketball players, the application of the methods of autogenic and ideomotor training according to our methodology. The study involved 63 students 1.5 courses from various universities of Kharkov 1.2 digits: 32 experimental group and 31 - control. The developed system of training students, basketball players used within 1 year. The efficiency of the developed system in the training process of students, basketball players.

  18. Shuttle TPS thermal performance and analysis methodology

    Science.gov (United States)

    Neuenschwander, W. E.; Mcbride, D. U.; Armour, G. A.

    1983-01-01

    Thermal performance of the thermal protection system was approximately as predicted. The only extensive anomalies were filler bar scorching and over-predictions in the high Delta p gap heating regions of the orbiter. A technique to predict filler bar scorching has been developed that can aid in defining a solution. Improvement in high Delta p gap heating methodology is still under study. Minor anomalies were also examined for improvements in modeling techniques and prediction capabilities. These include improved definition of low Delta p gap heating, an analytical model for inner mode line convection heat transfer, better modeling of structure, and inclusion of sneak heating. The limited number of problems related to penetration items that presented themselves during orbital flight tests were resolved expeditiously, and designs were changed and proved successful within the time frame of that program.

  19. Development of economic consequence methodology for process risk analysis.

    Science.gov (United States)

    Zadakbar, Omid; Khan, Faisal; Imtiaz, Syed

    2015-04-01

    A comprehensive methodology for economic consequence analysis with appropriate models for risk analysis of process systems is proposed. This methodology uses loss functions to relate process deviations in a given scenario to economic losses. It consists of four steps: definition of a scenario, identification of losses, quantification of losses, and integration of losses. In this methodology, the process deviations that contribute to a given accident scenario are identified and mapped to assess potential consequences. Losses are assessed with an appropriate loss function (revised Taguchi, modified inverted normal) for each type of loss. The total loss is quantified by integrating different loss functions. The proposed methodology has been examined on two industrial case studies. Implementation of this new economic consequence methodology in quantitative risk assessment will provide better understanding and quantification of risk. This will improve design, decision making, and risk management strategies.

  20. DOE 2009 Geothermal Risk Analysis: Methodology and Results (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Young, K. R.; Augustine, C.; Anderson, A.

    2010-02-01

    This presentation summarizes the methodology and results for a probabilistic risk analysis of research, development, and demonstration work-primarily for enhanced geothermal systems (EGS)-sponsored by the U.S. Department of Energy Geothermal Technologies Program.

  1. Integrated Methodology for Software Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Marian Pompiliu CRISTESCU

    2012-01-01

    Full Text Available The most used techniques to ensure safety and reliability of the systems are applied together as a whole, and in most cases, the software components are usually overlooked or to little analyzed. The present paper describes the applicability of fault trees analysis software system, analysis defined as Software Fault Tree Analysis (SFTA, fault trees are evaluated using binary decision diagrams, all of these being integrated and used with help from Java library reliability.

  2. Methodology for estimating radiation doses due to tritium and radiocarbon releases. [Health hazards from thermonuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D.A.; Soldat, J.K.

    1976-09-01

    Although the exact release rates of tritium (/sup 3/H) and /sup 14/C to the environment from a fusion power plant are not known, unit dose rates from postulated releases to air or to surface water can be calculated for a hypothetical individual and for population groups. Assuming a tritium release of 1 curie per year (Ci/yr) as HTO to the atmosphere, a hypothetical maximum individual residing near a fusion power plant might receive a dose rate of 2 x 10/sup -3/ millirem per year (mrem/yr). Assuming a 1 Ci/yr release to surface waters, this individual might receive a dose rate of 1.5 x 10/sup -5/ mrem/yr. The dose rate to the population of the world including the United Sates and the regional population was estimated to be 1 x 10/sup -2/ man-rem/yr from the release to the atmosphere and 6 x 10/sup -2/ man-rem/yr from the release to surface waters. Dose rates from releases of 1 Ci/yr /sup 14/C to the atmosphere were estimated to be 0.4 mrem/yr to the bone of the hypothetical maximum individual and 2 man-rem/yr to the total body of the world population. Because of the persistence of /sup 14/C in the environment and the fact that carbon is a major constituent of any living thing, efforts should be made to eliminate those releases with available technology such as double containment of the reactors to prevent air leakage.

  3. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  4. Design and analysis of multicavity prestressed concrete reactor vessels. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels.

  5. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  6. Radiochemical Analysis Methodology for uranium Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Scatena-Wachel DE

    2007-01-09

    This report provides sufficient material for a test sponsor with little or no radiochemistry background to understand and follow physics irradiation test program execution. Most irradiation test programs employ similar techniques and the general details provided here can be applied to the analysis of other irradiated sample types. Aspects of program management directly affecting analysis quality are also provided. This report is not an in-depth treatise on the vast field of radiochemical analysis techniques and related topics such as quality control. Instrumental technology is a very fast growing field and dramatic improvements are made each year, thus the instrumentation described in this report is no longer cutting edge technology. Much of the background material is still applicable and useful for the analysis of older experiments and also for subcontractors who still retain the older instrumentation.

  7. PROBABILISTIC METHODOLOGY OF LOW CYCLE FATIGUE ANALYSIS

    Institute of Scientific and Technical Information of China (English)

    Jin Hui; Wang Jinnuo; Wang Libin

    2003-01-01

    The cyclic stress-strain responses (CSSR), Neuber's rule (NR) and cyclic strain-life relation (CSLR) are treated as probabilistic curves in local stress and strain method of low cycle fatigue analy sis. The randomness of loading and the theory of fatigue damage accumulation (TOFDA) are consid ered. The probabilistic analysis of local stress, local strain and fatigue life are constructed based on the first-order Taylor's series expansions. Through this method proposed fatigue reliability analysis can be accomplished.

  8. Mass Spectrometry Methodology in Lipid Analysis

    OpenAIRE

    Lin Li; Juanjuan Han; Zhenpeng Wang; Jian'an Liu; Jinchao Wei; Shaoxiang Xiong; Zhenwen Zhao

    2014-01-01

    Lipidomics is an emerging field, where the structures, functions and dynamic changes of lipids in cells, tissues or body fluids are investigated. Due to the vital roles of lipids in human physiological and pathological processes, lipidomics is attracting more and more attentions. However, because of the diversity and complexity of lipids, lipid analysis is still full of challenges. The recent development of methods for lipid extraction and analysis and the combination with bioinformatics tech...

  9. SINGULAR SPECTRUM ANALYSIS: METHODOLOGY AND APPLICATION TO ECONOMICS DATA

    Institute of Scientific and Technical Information of China (English)

    Hossein HASSANI; Anatoly ZHIGLJAVSKY

    2009-01-01

    This paper describes the methodology of singular spectrum analysis (SSA) and demonstrate that it is a powerful method of time series analysis and forecasting, particulary for economic time series. The authors consider the application of SSA to the analysis and forecasting of the Iranian national accounts data as provided by the Central Bank of the Islamic Republic of lran.

  10. Finite element analysis applied to dentoalveolar trauma: methodology description.

    Science.gov (United States)

    da Silva, B R; Moreira Neto, J J S; da Silva, F I; de Aguiar, A S W

    2011-01-01

    Dentoalveolar traumatic injuries are among the clinical conditions most frequently treated in dental practice. However, few studies so far have addressed the biomechanical aspects of these events, probably as a result of difficulties in carrying out satisfactory experimental and clinical studies as well as the unavailability of truly scientific methodologies. The aim of this paper was to describe the use of finite element analysis applied to the biomechanical evaluation of dentoalveolar trauma. For didactic purposes, the methodological process was divided into steps that go from the creation of a geometric model to the evaluation of final results, always with a focus on methodological characteristics, advantages, and disadvantages, so as to allow the reader to customize the methodology according to specific needs. Our description shows that the finite element method can faithfully reproduce dentoalveolar trauma, provided the methodology is closely followed and thoroughly evaluated.

  11. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  12. Neutronic and thermal analysis of composite fuel for potential deployment in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abou Jaoude, Abdalla; Thomas, Colin; Erickson, Anna, E-mail: erickson@gatech.edu

    2016-07-15

    Highlights: • Neutronic and heat transfer performance of composite fuels on the macro-scale. • Methodology to guide flexible fuel design using high fidelity simulation tools. • Viability of composite fuels for ultra-high burnup fast reactor deployment. - Abstract: Composite fuels are promising candidates for high-burnup fast reactors because of their accommodation of swelling, limited fuel-cladding interactions and flexibility in design. While a proof-of-concept fuel consisting of granules of U-alloys and PuO{sub 2} dispersed within a porous zirconium matrix was successfully manufactured and irradiated, its neutronic and thermal performance remains to be optimized as compared to currently utilized fuels. MCNP6, COMSOL and a sphere packing algorithm were employed to perform the analysis. We found that both the theoretical maximum burnup reached and the temperature profiles are comparable to that of the currently considered alternative fuel. The results are promising and do not indicate any substantial limitation to the deployment of composite fuel. The fuel type merits further research, including full-core simulations. The methodology followed herein also provides a basis for screening different material compositions and guiding materials selection in composite fuels.

  13. A New Methodology of Spatial Crosscorrelation Analysis

    CERN Document Server

    Chen, Yanguang

    2015-01-01

    The idea of spatial crosscorrelation was conceived of long ago. However, unlike the related spatial autocorrelation, the theory and method of spatial crosscorrelation analysis have remained undeveloped. This paper presents a set of models and working methods for spatial crosscorrelation analysis. By analogy with Moran's index newly expressed in a spatial quadratic form and by means of mathematical reasoning, I derive a theoretical framework for geographical crosscorrelation analysis. First, two sets of spatial crosscorrelation coefficients are defined, including a global spatial crosscorrelation coefficient and a set of local spatial crosscorrelation coefficients. Second, a pair of scatterplots of spatial crosscorrelation is proposed, and different scatterplots show different relationships between correlated variables. Based on the spatial crosscorrelation coefficient, Pearson's correlation coefficient can be decomposed into two parts: direct correlation (partial crosscorrelation) and indirect correlation (sp...

  14. Stability analysis of a PFTR reactor for a first order kinetic reaction using the Lyapunov functionals

    Directory of Open Access Journals (Sweden)

    Héctor Armando Durán Peralta

    2010-04-01

    Full Text Available The stability of reactors having encompassing concentration and temperature parameters, such as continuous flow stirred tank reactors (CSTR, has been widely explored in the literature; however, there are few papers about the stability of tubular reactor having distributed spatial concentration and temperature parameters such as the plow flow tubular reactor (PFTR. This paper analyses the stability of isothermal and non-isothermal PFTR reactors using the Lyapunov functional method. The first order kinetic reaction was selected because one of this paper’s oblectives was to apply Lyapunov functionals to stability analysis of distributed parameter reactors (technique used in electrical engineering systems’ stability analysis. The stability analysis revealed asymptotically stable tempe- rature and concentration profiles for isothermal PFTR, non-isothermal PFTR with kinetic constant independent of temperature and adiabatic non-isothermal PFTR. Analysis revealed an asymptotically stability region for the heat exchange reactor and an uncertain region where it may have oscillations.

  15. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)

    2014-05-15

    The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.

  16. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  17. Analysis of Alternatives for Risk Assessment Methodologies and Tools

    Energy Technology Data Exchange (ETDEWEB)

    Nachtigal, Noel M. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). System Analytics; Fruetel, Julia A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Gleason, Nathaniel J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Helms, Jovana [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Imbro, Dennis Raymond [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Sumner, Matthew C. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis

    2013-10-01

    The purpose of this document is to provide a basic overview and understanding of risk assessment methodologies and tools from the literature and to assess the suitability of these methodologies and tools for cyber risk assessment. Sandia National Laboratories (SNL) performed this review in support of risk modeling activities performed for the Stakeholder Engagement and Cyber Infrastructure Resilience (SECIR) division of the Department of Homeland Security (DHS) Office of Cybersecurity and Communications (CS&C). The set of methodologies and tools covered in this document is not intended to be exhaustive; instead, it focuses on those that are commonly used in the risk assessment community. The classification of methodologies and tools was performed by a group of analysts with experience in risk analysis and cybersecurity, and the resulting analysis of alternatives has been tailored to address the needs of a cyber risk assessment.

  18. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  19. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  20. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  1. Analysis and Evaluation Methodology of Effect of Light Water Reactor Coolant Environment on Fatigue Life of Class 1 Components%轻水堆冷却剂环境对核一级部件疲劳寿命影响的分析与评价方法

    Institute of Scientific and Technical Information of China (English)

    房永刚; 王庆; 初起宝; 张跃; 孙造占; 王臣

    2013-01-01

    It is required that the fatigue analysis should be done for nuclear reactor class 1 components composing of the reactor pressure boundary .Based on worldwide resear-ches ,it is found that the fatigue design curves currently used in the analysis have no enough margins considering the coolant environmental effect ,and the regulatory bodies in the world have paid high attention to it .General information related to the research background was introduced ,and the NRC method and the related ASME activities were discussed .Possible ways are recommended to deal with the potential issues that may arise .%反应堆压力边界的核一级部件在设计中要求进行疲劳分析。当前国际上的研究结果表明,目前分析中使用的规范疲劳设计曲线在考虑冷却剂环境条件下并不保守,并引起了各国核电监管机构对此的广泛关注。文中介绍了各主要核电国家对冷却剂环境疲劳的研究情况,讨论了N RC关于冷却剂环境的疲劳分析方法以及ASM E规范的后续进展。针对考虑环境疲劳后可能带来的一系列问题,提出了建议的解决方法。

  2. Toward a standardized urine proteome analysis methodology.

    Science.gov (United States)

    Court, Magali; Selevsek, Nathalie; Matondo, Mariette; Allory, Yves; Garin, Jerome; Masselon, Christophe D; Domon, Bruno

    2011-03-01

    Urine is an easily accessible bodily fluid particularly suited for the routine clinical analysis of disease biomarkers. Actually, the urinary proteome is more diverse than anticipated a decade ago. Hence, significant analytical and practical issues of urine proteomics such as sample collection and preparation have emerged, in particular for large-scale studies. We have undertaken a systematic study to define standardized and integrated analytical protocols for a biomarker development pipeline, employing two LC-MS analytical platforms, namely accurate mass and time tags and selected reaction monitoring, for the discovery and verification phase, respectively. Urine samples collected from hospital patients were processed using four different protocols, which were evaluated and compared on both analytical platforms. Addition of internal standards at various stages of sample processing allowed the estimation of protein extraction yields and the absolute quantification of selected urinary proteins. Reproducibility of the entire process and dynamic range of quantification were also evaluated. Organic solvent precipitation followed by in-solution digestion provided the best performances and was thus selected as the standard method common to the discovery and verification phases. Finally, we applied this protocol for platforms' cross-validation and obtained excellent consistency between urinary protein concentration estimates by both analytical methods performed in parallel in two laboratories. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  3. Methodology for reliability allocation based on fault tree analysis and dualistic contrast

    Institute of Scientific and Technical Information of China (English)

    TONG Lili; CAO Xuewu

    2008-01-01

    Reliability allocation is a difficult multi-objective optimization problem.This paper presents a methodology for reliability allocation that can be applied to determine the reliability characteristics of reactor systems or subsystems.The dualistic contrast,known as one of the most powerful tools for optimization problems,is applied to the reliability allocation model of a typical system in this article.And the fault tree analysis,deemed to be one of the effective methods of reliability analysis,is also adopted.Thus a failure rate allocation model based on the fault tree analysis and dualistic contrast is achieved.An application on the emergency diesel generator in the nuclear power plant is given to illustrate the proposed method.

  4. Advanced Power Plant Development and Analysis Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    A.D. Rao; G.S. Samuelsen; F.L. Robson; B. Washom; S.G. Berenyi

    2006-06-30

    Under the sponsorship of the U.S. Department of Energy/National Energy Technology Laboratory, a multi-disciplinary team led by the Advanced Power and Energy Program of the University of California at Irvine is defining the system engineering issues associated with the integration of key components and subsystems into advanced power plant systems with goals of achieving high efficiency and minimized environmental impact while using fossil fuels. These power plant concepts include 'Zero Emission' power plants and the 'FutureGen' H2 co-production facilities. The study is broken down into three phases. Phase 1 of this study consisted of utilizing advanced technologies that are expected to be available in the 'Vision 21' time frame such as mega scale fuel cell based hybrids. Phase 2 includes current state-of-the-art technologies and those expected to be deployed in the nearer term such as advanced gas turbines and high temperature membranes for separating gas species and advanced gasifier concepts. Phase 3 includes identification of gas turbine based cycles and engine configurations suitable to coal-based gasification applications and the conceptualization of the balance of plant technology, heat integration, and the bottoming cycle for analysis in a future study. Also included in Phase 3 is the task of acquiring/providing turbo-machinery in order to gather turbo-charger performance data that may be used to verify simulation models as well as establishing system design constraints. The results of these various investigations will serve as a guide for the U. S. Department of Energy in identifying the research areas and technologies that warrant further support.

  5. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  6. Refined multi-level methodology in parallel computing environment for BWR RIA analysis

    Science.gov (United States)

    Solis-Rodarte, Jorge

    2000-12-01

    Best-estimate methodologies in boiling water reactor can reduce the traditional conservative thermal margins imposed on the designs and during the operation of this type of nuclear reactors. Traditional operating thermal margins are obtained based on simplified modeling techniques that are supplemented with the required dose of conservatism. For instance, much more realistic transient pin peaking distributions can be predicted by applying a dehomogenization algorithm, based on a flux reconstruction scheme which uses nodal results during both steady state and transient calculation at each time step. A subchannel analysis module for obtaining thermal margins supplements the calculation approach used. A multi-level methodology to extend the TRAC-BF1/NEM coupled code capability to obtain the transient fuel rod response has been implemented. To fulfill the development requirements some improved neutronic models were implemented into the NEM solution algorithm, namely the pin power reconstruction capability, and the simulation of a dynamic scram. The obtained results were coupled to a subchannel analysis module: COBRA-TF T-H subchannel analysis code. The objective of the pin power reconstruction capability of NEM is to locate the most limiting node (axial region of assembly/channel) within the core. The power information obtained from the NEM 3D neutronic calculation is used by the hot channel analysis module (COBRA-TF). COBRA-TF needs also the T-H conditions at the boundary nodes. This information is provided by TRACBF1 T-H system analysis code. The Subchannel analysis module uses this information to re-calculate the fluid, thermal and hydraulics conditions in the most limiting node (axial region of assembly/channel) within the core.

  7. Dynamic analysis on methanation reactor using a double-input-multi-output linearized model☆

    Institute of Scientific and Technical Information of China (English)

    Xingxing Li; Jiageng Li; Bolun Yang; Yong Zhang

    2015-01-01

    A double-input–multi-output linearized system is developed using the state-space method for dynamic analysis of methanation process of coke oven gas. The stability of reactor alone and reactor with feed-effluent heat ex-changer is compared through the dominant poles of the system transfer functions. With single or double distur-bance of temperature and CO concentration at the reactor inlet, typical dynamic behavior in the reactor, including fast concentration response, slow temperature response and inverse response, is revealed for further under-standing of the counteraction and synergy effects caused by simultaneous variation of concentration and temper-ature. Analysis results show that the stability of the reactor loop is more sensitive than that of reactor alone due to the positive heat feedback. Remarkably, with the decrease of heat exchange efficiency, the reactor system may display limit cycle behavior for a pair of complex conjugate poles across the imaginary axis.

  8. CONTENT ANALYSIS, DISCOURSE ANALYSIS, AND CONVERSATION ANALYSIS: PRELIMINARY STUDY ON CONCEPTUAL AND THEORETICAL METHODOLOGICAL DIFFERENCES

    Directory of Open Access Journals (Sweden)

    Anderson Tiago Peixoto Gonçalves

    2016-08-01

    Full Text Available This theoretical essay aims to reflect on three models of text interpretation used in qualitative research, which is often confused in its concepts and methodologies (Content Analysis, Discourse Analysis, and Conversation Analysis. After the presentation of the concepts, the essay proposes a preliminary discussion on conceptual and theoretical methodological differences perceived between them. A review of the literature was performed to support the conceptual and theoretical methodological discussion. It could be verified that the models have differences related to the type of strategy used in the treatment of texts, the type of approach, and the appropriate theoretical position.

  9. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Matjaž Leskovar

    2016-02-01

    Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

  10. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  11. An Economic Analysis of Generation IV Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, J S; Lamont, A D; Rothwell, G S; Smith, C F; Greenspan, E; Brown, N; Barak, A

    2002-03-01

    This report examines some conditions necessary for Generation IV Small Modular Reactors (SMRs) to be competitive in the world energy market. The key areas that make nuclear reactors an attractive choice for investors are reviewed, and a cost model based on the ideal conditions is developed. Recommendations are then made based on the output of the cost model and on conditions and tactics that have proven successful in other industries. The Encapsulated Nuclear Heat Source (ENHS), a specific SMR design concept, is used to develop the cost model and complete the analysis because information about the ENHS design is readily available from the University of California at Berkeley Nuclear Engineering Department. However, the cost model can be used to analyze any of the current SMR designs being considered. On the basis of our analysis, we determined that the nuclear power industry can benefit from and SMRs can become competitive in the world energy market if a combination of standardization and simplification of orders, configuration, and production are implemented. This would require wholesale changes in the way SMRs are produced, manufactured and regulated, but nothing that other industries have not implemented and proven successful.

  12. Full core analysis of IRIS reactor by using MCNPX.

    Science.gov (United States)

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.

  13. Job task and functional analysis of the Division of Reactor Projects, office of Nuclear Reactor Regulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Morzinski, J.A.; Gilmore, W.; Hahn, H.A.

    1998-07-10

    A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.

  14. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  15. A Global Sensitivity Analysis Methodology for Multi-physics Applications

    Energy Technology Data Exchange (ETDEWEB)

    Tong, C H; Graziani, F R

    2007-02-02

    Experiments are conducted to draw inferences about an entire ensemble based on a selected number of observations. This applies to both physical experiments as well as computer experiments, the latter of which are performed by running the simulation models at different input configurations and analyzing the output responses. Computer experiments are instrumental in enabling model analyses such as uncertainty quantification and sensitivity analysis. This report focuses on a global sensitivity analysis methodology that relies on a divide-and-conquer strategy and uses intelligent computer experiments. The objective is to assess qualitatively and/or quantitatively how the variabilities of simulation output responses can be accounted for by input variabilities. We address global sensitivity analysis in three aspects: methodology, sampling/analysis strategies, and an implementation framework. The methodology consists of three major steps: (1) construct credible input ranges; (2) perform a parameter screening study; and (3) perform a quantitative sensitivity analysis on a reduced set of parameters. Once identified, research effort should be directed to the most sensitive parameters to reduce their uncertainty bounds. This process is repeated with tightened uncertainty bounds for the sensitive parameters until the output uncertainties become acceptable. To accommodate the needs of multi-physics application, this methodology should be recursively applied to individual physics modules. The methodology is also distinguished by an efficient technique for computing parameter interactions. Details for each step will be given using simple examples. Numerical results on large scale multi-physics applications will be available in another report. Computational techniques targeted for this methodology have been implemented in a software package called PSUADE.

  16. An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

    2015-02-15

    Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

  17. Vectorized Monte Carlo methods for reactor lattice analysis

    Science.gov (United States)

    Brown, F. B.

    1984-01-01

    Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

  18. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  19. PRODUCTION OF MEDIUM-CHAIN ACYLGLYCEROLS BY LIPASE ESTERIFICATION IN PACKED BED REACTOR: PROCESS OPTIMIZATION BY RESPONSE SURFACE METHODOLOGY

    Directory of Open Access Journals (Sweden)

    ZANARIAH MOHD DOM

    2014-06-01

    Full Text Available Medium-chain acylglycerols (or glycerides are formed of mono-, di- and triacylglycerol classes. In this study, an alternative method to produce MCA from esterifying palm oil fatty acid distillate (PFAD with the presence of oil palm mesocarp lipase (OPML which is a plant-sourced lipase and PFAD is also cheap by-product is developed in a packed bed reactor. The production of medium-chain acylglycerols (MCA by lipase-catalysed esterification of palm oil fatty acid distillate with glycerol are optimize in order to determine the factors that have significant effects on the reaction condition and high yield of MCA. Response surface methodology (RSM was applied to optimize the reaction conditions. The reaction conditions, namely, the reaction time (30-240 min, enzyme load (0.5-1.5 kg, silica gel load (0.2-1.0 kg, and solvent amount (200-600 vol/wt. Reaction time, enzyme loading and solvent amount strongly effect MCA synthesis (p0.05 influence on MCA yield. Best-fitting models were successfully established for MCA yield (R 2 =0.9133. The optimum MCA yield were 75% from the predicted value and 75.4% from the experimental data for 6 kg enzyme loading, a reaction time of 135min and a solvent amount of 350 vol/wt at 65ºC reaction temperature. Verification of experimental results under optimized reaction conditions were conducted, and the results agreed well with the predicted range. Esterification products (mono-, di- and triacylglycerol from the PBR were identified using Thin Layer Chromatography method. The chromatograms showed the successful fractionation of esterified products in this alternative method of process esterification.

  20. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  1. Multiphysics methods development for high temperature gas reactor analysis

    Science.gov (United States)

    Seker, Volkan

    Multiphysics computational methods were developed to perform design and safety analysis of the next generation Pebble Bed High Temperature Gas Cooled Reactors. A suite of code modules was developed to solve the coupled thermal-hydraulics and neutronics field equations. The thermal-hydraulics module is based on the three dimensional solution of the mass, momentum and energy equations in cylindrical coordinates within the framework of the porous media method. The neutronics module is a part of the PARCS (Purdue Advanced Reactor Core Simulator) code and provides a fine mesh finite difference solution of the neutron diffusion equation in three dimensional cylindrical coordinates. Coupling of the two modules was performed by mapping the solution variables from one module to the other. Mapping is performed automatically in the code system by the use of a common material mesh in both modules. The standalone validation of the thermal-hydraulics module was performed with several cases of the SANA experiment and the standalone thermal-hydraulics exercise of the PBMR-400 benchmark problem. The standalone neutronics module was validated by performing the relevant exercises of the PBMR-268 and PBMR-400 benchmark problems. Additionally, the validation of the coupled code system was performed by analyzing several steady state and transient cases of the OECD/NEA PBMR-400 benchmark problem.

  2. Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

    1990-11-01

    A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

  3. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  4. Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code

    Science.gov (United States)

    Tiyapun, K.; Wetchagarun, S.

    2017-06-01

    The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.

  5. Active disturbance rejection control: methodology and theoretical analysis.

    Science.gov (United States)

    Huang, Yi; Xue, Wenchao

    2014-07-01

    The methodology of ADRC and the progress of its theoretical analysis are reviewed in the paper. Several breakthroughs for control of nonlinear uncertain systems, made possible by ADRC, are discussed. The key in employing ADRC, which is to accurately determine the "total disturbance" that affects the output of the system, is illuminated. The latest results in theoretical analysis of the ADRC-based control systems are introduced.

  6. Architecture dependent availability analysis of RPS for Research Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.

  7. Thermal stability analysis of the liquid phase methanol synthesis reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gogate, M.R.; Desirazu, S.; Berty, J.M.; Lee, S. (Akron University, Akron, OH (USA). Dept. of Chemical Engineering)

    1992-01-01

    The effect of addition of an inert liquid phase on the rate of heat generation in the catalytic synthesis of methanol from syngas has been studied. Gas compositions typical of product gases from Lurgi and Koppers-Totzek gasifiers, represented by H[sub 2]-rich and CO-rich syngas respectively, were used to experimentally verify the 'slope' and 'dynamic' criteria in a three-phase fixed bed recycle reactor. The liquid medium, Witco-40 oil, has been effective in controlling the rate of heat generation and in preventing catalyst overheating, signifying that the liquid phase synthesis is thermally far more stable than the vapour phase synthesis. The experimental thermal stability study provides crucial and valuable information in commercializing the liquid phase methanol synthesis process. The current approach of thermal stability analysis does not require any a priori assumption or predetermined reaction kinetics. 22 refs., 6 figs., 7 tabs.

  8. Methodology for dimensional variation analysis of ITER integrated systems

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes, F. Javier, E-mail: FranciscoJavier.Fuentes@iter.org [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France); Trouvé, Vincent [Assystem Engineering & Operation Services, rue J-M Jacquard CS 60117, 84120 Pertuis (France); Cordier, Jean-Jacques; Reich, Jens [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  9. Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor; Analyses de sensibilite et d'incertitude de donnees nucleaires. Contribution a la validation d'une methodologie utilisant la theorie des perturbations; application a un concept innovant: reacteur a sels fondus thorium a spectre epithermique

    Energy Technology Data Exchange (ETDEWEB)

    Bidaud, A

    2005-10-15

    Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

  10. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1993-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor (AFR) criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial pressurized-water reactors (PWR). The analysis methodology selected for all calculations reported herein was the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted comparison of criticality calculations directly using the utility-calculated isotopics to those using the isotopics generated by the SCALE-4 SAS2H

  11. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  12. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  13. The application of research reactor Maria for analysis of thorium use in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)

    2010-07-01

    The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)

  14. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  15. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  16. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  17. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  18. Continuous biosynthesis of biodiesel from waste cooking palm oil in a packed bed reactor: optimization using response surface methodology (RSM) and mass transfer studies.

    Science.gov (United States)

    Halim, Siti Fatimah Abdul; Kamaruddin, Azlina Harun; Fernando, W J N

    2009-01-01

    This study aimed to develop an optimal continuous procedure of lipase-catalyzes transesterification of waste cooking palm oil in a packed bed reactor to investigate the possibility of large scale production further. Response surface methodology (RSM) based on central composite rotatable design (CCRD) was used to optimize the two important reaction variables packed bed height (cm) and substrate flow rate(ml/min) for the transesterification of waste cooking palm oil in a continuous packed bed reactor. The optimum condition for the transesterification of waste cooking palm oil was as follows: 10.53 cm packed bed height and 0.57 ml/min substrate flow rate. The optimum predicted fatty acid methyl ester (FAME) yield was 80.3% and the actual value was 79%. The above results shows that the RSM study based on CCRD is adaptable for FAME yield studied for the current transesterification system. The effect of mass transfer in the packed bed reactor has also been studied. Models for FAME yield have been developed for cases of reaction control and mass transfer control. The results showed very good agreement compatibility between mass transfer model and the experimental results obtained from immobilized lipase packed bed reactor operation, showing that in this case the FAME yield was mass transfer controlled.

  19. A Case Study on Methodological Approaches to Conversation Analysis

    Institute of Scientific and Technical Information of China (English)

    廉莲

    2014-01-01

    Taking a piece of social interaction as the object of the study, some basic and brief analysis on how meaning is negotiat-ed is offered from both structural and functional perspectives. The potential purpose is to provide readers with a maybe rough but clear presentation of those assorted methods used in conversation analysis. Out of the presentation is developed a possibility for language learners as well as teachers to be more aware of the differences and also the interrelations among these methodological approaches to conversation analysis, which may be of some relevance to teaching practice.

  20. Markovian reliability analysis under uncertainty with an application on the shutdown system of the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Papazoglou, I A; Gyftopoulos, E P

    1978-09-01

    A methodology for the assessment of the uncertainties about the reliability of nuclear reactor systems described by Markov models is developed, and the uncertainties about the probability of loss of coolable core geometry (LCG) of the Clinch River Breeder Reactor (CRBR) due to shutdown system failures, are assessed. Uncertainties are expressed by assuming the failure rates, the repair rates and all other input variables of reliability analysis as random variables, distributed according to known probability density functions (pdf). The pdf of the reliability is then calculated by the moment matching technique. Two methods have been employed for the determination of the moments of the reliability: the Monte Carlo simulation; and the Taylor-series expansion. These methods are adopted to Markovian problems and compared for accuracy and efficiency.

  1. Modified divergence theorem for analysis and optimization of wall reflecting cylindrical UV reactor

    Directory of Open Access Journals (Sweden)

    Milanović Đurđe R.

    2011-01-01

    Full Text Available In this paper, Modified Divergence Theorem (MDT, known in earlier literature as Gauss-Ostrogradsky theorem, was formulated and proposed as a general approach to electromagnetic (EM radiation, especially ultraviolet (UV radiation reactor modeling. Formulated mathematical model, based on MDT, for multilamp UV reactor was applied to all sources in a reactor in order to obtain intensity profiles at chosen surfaces inside reactor. Applied modification of MDT means that intensity at a real opaque or transparent surface or through a virtual surface, opened or closed, from different sides of the surface are added and not subtracted as in some other areas of physics. Derived model is applied to an example of the multiple UV sources reactor, where sources are arranged inside a cylindrical reactor at the coaxial virtual cylinder, having the radius smaller than the radius of the reactor. In this work, optimization of a reactor means maximum transfer of EM energy sources into the fluid for given fluid absorbance and fluid flow-dose product. Obtained results, for in advanced known water quality, gives unique solution for an optimized model of a multilamp reactor geometry. As everyone can easily verify, MDT is very good starting point for every reactor modeling and analysis.

  2. A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author).

  3. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  4. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    Science.gov (United States)

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  5. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    Science.gov (United States)

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  6. Evaluation on safety concerns of integral reactor: development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Lee, S. G.; Sin, A. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They includes the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. These efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. 62 refs., 3 figs., 21 tabs. (Author)

  7. Development of analysis system and analysis on reactor physics for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Gi; Bae, Chang Joon; Kwon, Oh Sun [Korea Power Electric Corporation, Taejon (Korea, Republic of)

    1997-07-01

    characteristics of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU= fuel core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. (Author) 32 refs., 25 tabs., 79 figs.

  8. Transfinite element methodology towards a unified thermal/structural analysis

    Science.gov (United States)

    Tamma, K. K.; Railkar, S. B.

    1986-01-01

    The paper describes computational developments towards thermal/structural modeling and analysis via a generalized common numerical methodology for effectively and efficiently interfacing interdisciplinary areas. The proposed formulations use transform methods in conjunction with finite element developments for each of the heat transfer and structural disciplines, respectively, providing avenues for obtaining the structural response due to thermal effects. An alternative methodology for unified thermal/structural analysis is presented. The potential of the approach is outlined in comparison with conventional schemes and existing practices. Highlights and characteristic features of the approach are described via general formulations and applications to several problems. Results obtained demonstrate excellent agreement in comparison with analytic and/or conventional finite element schemes accurately and efficiently.

  9. Latest developments on safety analysis methodologies at the Juzbado plant

    Energy Technology Data Exchange (ETDEWEB)

    Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A. [ENUSA Industrias Avanzadas S. A., Juzbado Nuclear Fuel Fabrication Plant, Ctra. Salamanca-Ledesma, km. 26, 37015 Juzbado, Salamanca (Spain)

    2010-07-01

    Over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. This paper describes the three most outstanding of them, so as to say, the Integrated Safety Analysis (ISA) project, the adaptation of the MARSSIM methodology for characterization surveys of radioactive contamination spots, and the programme for the Systematic Review of the Operational Conditions of the Safety Systems (SROCSS). Several reasons motivated the decision to implement such methodologies, such as Regulator requirements, operational experience and of course, the strong commitment of ENUSA to maintain the highest standards of nuclear industry on all the safety relevant activities. In this context, since 2004 ENUSA is undertaking the ISA project, which consists on a systematic examination of plant's processes, equipment, structures and personnel activities to ensure that all relevant hazards that could result in unacceptable consequences have been adequately evaluated and the appropriate protective measures have been identified. On the other hand and within the framework of a current programme to ensure the absence of radioactive contamination spots on unintended areas, the MARSSIM methodology is being applied as a tool to conduct the radiation surveys and investigation of potentially contaminated areas. Finally, the SROCSS programme was initiated earlier this year 2009 to assess the actual operating conditions of all the systems with safety relevance, aiming to identify either potential non-conformities or areas for improvement in order to ensure their high performance after years of operation. The following paragraphs describe the key points related to these three methodologies as well as an outline of the results obtained so far. (authors)

  10. Event-scale power law recession analysis: quantifying methodological uncertainty

    Science.gov (United States)

    Dralle, David N.; Karst, Nathaniel J.; Charalampous, Kyriakos; Veenstra, Andrew; Thompson, Sally E.

    2017-01-01

    The study of single streamflow recession events is receiving increasing attention following the presentation of novel theoretical explanations for the emergence of power law forms of the recession relationship, and drivers of its variability. Individually characterizing streamflow recessions often involves describing the similarities and differences between model parameters fitted to each recession time series. Significant methodological sensitivity has been identified in the fitting and parameterization of models that describe populations of many recessions, but the dependence of estimated model parameters on methodological choices has not been evaluated for event-by-event forms of analysis. Here, we use daily streamflow data from 16 catchments in northern California and southern Oregon to investigate how combinations of commonly used streamflow recession definitions and fitting techniques impact parameter estimates of a widely used power law recession model. Results are relevant to watersheds that are relatively steep, forested, and rain-dominated. The highly seasonal mediterranean climate of northern California and southern Oregon ensures study catchments explore a wide range of recession behaviors and wetness states, ideal for a sensitivity analysis. In such catchments, we show the following: (i) methodological decisions, including ones that have received little attention in the literature, can impact parameter value estimates and model goodness of fit; (ii) the central tendencies of event-scale recession parameter probability distributions are largely robust to methodological choices, in the sense that differing methods rank catchments similarly according to the medians of these distributions; (iii) recession parameter distributions are method-dependent, but roughly catchment-independent, such that changing the choices made about a particular method affects a given parameter in similar ways across most catchments; and (iv) the observed correlative relationship

  11. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  12. Comparative microbial analysis before and after foaming incidents in biogas reactors

    DEFF Research Database (Denmark)

    Kougias, Panagiotis; De Francisci, Davide; Treu, Laura

    2014-01-01

    retention time (HRT) of all reactors was kept constant at 15 days. The whole experiment was divided into two periods. During the first period, the reactors were fed only with cattle manure. Once steady state conditions were reached, liquid sample from all reactors was obtained for DNA extraction...... foam was steady, samples were taken again for DNA extraction and metagenomic analysis. Results from the present study revealed significant variations in the microbiology of the manure-based biogas reactors after foam initiation. A number of genera could be linked to foaming as they produce...

  13. Finite Element Analysis Applied to Dentoalveolar Trauma: Methodology Description

    OpenAIRE

    2011-01-01

    Dentoalveolar traumatic injuries are among the clinical conditions most frequently treated in dental practice. However, few studies so far have addressed the biomechanical aspects of these events, probably as a result of difficulties in carrying out satisfactory experimental and clinical studies as well as the unavailability of truly scientific methodologies. The aim of this paper was to describe the use of finite element analysis applied to the biomechanical evaluation of dentoalveolar traum...

  14. Uncertainty analysis of fission fraction for reactor antineutrino experiments

    Science.gov (United States)

    Ma, X. B.; Lu, F.; Wang, L. Z.; Chen, Y. X.; Zhong, W. L.; An, F. P.

    2016-06-01

    Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Therefore, how to evaluate the antineutrino flux uncertainty results from reactor simulation is an important question. In this study, a method of the antineutrino flux uncertainty result from reactor simulation was proposed by considering the correlation coefficient. In order to use this method in the Daya Bay antineutrino experiment, the open source code DRAGON was improved and used for obtaining the fission fraction and correlation coefficient. The average fission fraction between DRAGON and SCIENCE code was compared and the difference was less than 5% for all the four isotopes. The uncertainty of fission fraction was evaluated by comparing simulation atomic density of four main isotopes with Takahama-3 experiment measurement. After that, the uncertainty of the antineutrino flux results from reactor simulation was evaluated as 0.6% per core for Daya Bay antineutrino experiment.

  15. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  16. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  17. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  18. Energy efficiency analysis of reactor for torrefaction of biomass with direct heating

    Science.gov (United States)

    Kuzmina, J. S.; Director, L. B.; Shevchenko, A. L.; Zaichenko, V. M.

    2016-11-01

    Paper presents energy analysis of reactor for torrefaction with direct heating of granulated biomass by exhaust gases. Various schemes of gas flow through the reactor zones are presented. Performed is a comparative evaluation of the specific energy consumption for the considered schemes. It has been shown that one of the most expensive processes of torrefaction technology is recycling of pyrolysis gases.

  19. Degraded core analysis for the pressurized-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-02-09

    An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.

  20. 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.

  1. Development of a methodology for simulation of gas cooled reactors with purpose of transmutation; Desenvolvimento de uma metodologia para simulacao de reatores refrigerados a gas com proposito de transmutacao

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson Alberto da

    2009-07-01

    This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k{sub eff}, neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

  2. Simplifying multivariate survival analysis using global score test methodology

    Science.gov (United States)

    Zain, Zakiyah; Aziz, Nazrina; Ahmad, Yuhaniz

    2015-12-01

    In clinical trials, the main purpose is often to compare efficacy between experimental and control treatments. Treatment comparisons often involve multiple endpoints, and this situation further complicates the analysis of survival data. In the case of tumor patients, endpoints concerning survival times include: times from tumor removal until the first, the second and the third tumor recurrences, and time to death. For each patient, these endpoints are correlated, and the estimation of the correlation between two score statistics is fundamental in derivation of overall treatment advantage. In this paper, the bivariate survival analysis method using the global score test methodology is extended to multivariate setting.

  3. PANSYSTEMS ANALYSIS: MATHEMATICS, METHODOLOGY,RELATIVITY AND DIALECTICAL THINKING

    Institute of Scientific and Technical Information of China (English)

    郭定和; 吴学谋; 冯向军; 李永礼

    2001-01-01

    Based on new analysis modes and new definitions with relative mathematization and simplification or strengthening forms for concepts of generalized systems,panderivatives , pansymmetry , panbox principle, pansystems relativity, etc. , the framework and related principles of pansystems methodology and pansystems relativity are developed. Related contents include: pansystems with relatively universal mathematizing forns, 200 types of dualities, duality transformation, pansymmetry transformation,pansystems dialectics, the 8-domain method, pansystems mathematical methods,generalized quantification, the principles of approximation-transforming, pan-equivalence theorems , supply-demand analysis, thinking experiment, generalized gray systems, etc.

  4. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  5. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  6. Development of test methodology for dynamic mechanical analysis instrumentation

    Science.gov (United States)

    Allen, V. R.

    1982-08-01

    Dynamic mechanical analysis instrumentation was used for the development of specific test methodology in the determination of engineering parameters of selected materials, esp. plastics and elastomers, over a broad range of temperature with selected environment. The methodology for routine procedures was established with specific attention given to sample geometry, sample size, and mounting techniques. The basic software of the duPont 1090 thermal analyzer was used for data reduction which simplify the theoretical interpretation. Clamps were developed which allowed 'relative' damping during the cure cycle to be measured for the fiber-glass supported resin. The correlation of fracture energy 'toughness' (or impact strength) with the low temperature (glassy) relaxation responses for a 'rubber-modified' epoxy system was negative in result because the low-temperature dispersion mode (-80 C) of the modifier coincided with that of the epoxy matrix, making quantitative comparison unrealistic.

  7. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

  8. Methodological challenges in qualitative content analysis: A discussion paper.

    Science.gov (United States)

    Graneheim, Ulla H; Lindgren, Britt-Marie; Lundman, Berit

    2017-09-01

    This discussion paper is aimed to map content analysis in the qualitative paradigm and explore common methodological challenges. We discuss phenomenological descriptions of manifest content and hermeneutical interpretations of latent content. We demonstrate inductive, deductive, and abductive approaches to qualitative content analysis, and elaborate on the level of abstraction and degree of interpretation used in constructing categories, descriptive themes, and themes of meaning. With increased abstraction and interpretation comes an increased challenge to demonstrate the credibility and authenticity of the analysis. A key issue is to show the logic in how categories and themes are abstracted, interpreted, and connected to the aim and to each other. Qualitative content analysis is an autonomous method and can be used at varying levels of abstraction and interpretation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  10. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  11. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  12. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

  13. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  14. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  15. EVALUATION OF ACTIVATION PRODUCTS IN REMAINING IN REMAINING K-, L- AND C-REACTOR STRUCTURES

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.; Webb, R.

    2010-09-30

    An analytic model and calculational methodology was previously developed for P-reactor and R-reactor to quantify the radioisotopes present in Savannah River Site (SRS) reactor tanks and the surrounding structural materials as a result of neutron activation of the materials during reactor operation. That methodology has been extended to K-reactor, L-reactor, and C-reactor. The analysis was performed to provide a best-estimate source term input to the Performance Assessment for an in-situ disposition strategy by Site Decommissioning and Demolition (SDD). The reactor structure model developed earlier for the P-reactor and R-reactor analyses was also used for the K-reactor and L-reactor. The model was suitably modified to handle the larger Creactor tank and associated structures. For all reactors, the structure model consisted of 3 annular zones, homogenized by the amount of structural materials in the zone, and 5 horizontal layers. The curie content on an individual radioisotope basis and total basis for each of the regions was determined. A summary of these results are provided herein. The efficacy of this methodology to accurately predict the radioisotopic content of the reactor systems in question has been demonstrated and is documented in Reference 1. As noted in that report, results for one reactor facility cannot be directly extrapolated to other SRS reactors.

  16. A study on naphtha catalytic reforming reactor simulation and analysis

    OpenAIRE

    Liang, Ke-min; Guo, Hai-Yan; Pan, Shi-Wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation uni...

  17. A study on naphtha catalytic reforming reactor simulation and analysis.

    Science.gov (United States)

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-06-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  18. Study and Analysis on Naphtha Catalytic Reforming Reactor Simulation

    Institute of Scientific and Technical Information of China (English)

    Liang Ke min; Song Yongji; Pan Shiwei

    2004-01-01

    A naphtha catalytic reforming unit with four reactors connected in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reaction characteristics based on idealizing the complex naphtha mixture to represent the paraffin, naphthene, and aromatic groups with individual compounds. The simulation results based on above models agree very well with actual operating data of process unit.

  19. A study on naphtha catalytic reforming reactor simulation and analysis

    Institute of Scientific and Technical Information of China (English)

    LIANG Ke-min; GUO Hai-yan; PAN Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  20. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  1. Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70). 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. Note 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity). 1.3 This practice is restri...

  2. 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

  3. Evaluating some Reliability Analysis Methodologies in Seismic Design

    Directory of Open Access Journals (Sweden)

    A. E. Ghoulbzouri

    2011-01-01

    Full Text Available Problem statement: Accounting for uncertainties that are present in geometric and material data of reinforced concrete buildings is performed in this study within the context of performance based seismic engineering design. Approach: Reliability of the expected performance state is assessed by using various methodologies based on finite element nonlinear static pushover analysis and specialized reliability software package. Reliability approaches that were considered included full coupling with an external finite element code and surface response based methods in conjunction with either first order reliability method or importance sampling method. Various types of probability distribution functions that model parameters uncertainties were introduced. Results: The probability of failure according to the used reliability analysis method and to the selected distribution of probabilities was obtained. Convergence analysis of the importance sampling method was performed. The required duration of analysis as function of the used reliability method was evaluated. Conclusion/Recommendations: It was found that reliability results are sensitive to the used reliability analysis method and to the selected distribution of probabilities. Durations of analysis for coupling methods were found to be higher than those associated to surface response based methods; one should however include time needed to derive these lasts. For the reinforced concrete building considered in this study, it was found that significant variations exist between all the considered reliability methodologies. The full coupled importance sampling method is recommended, but the first order reliability method applied on a surface response model can be used with good accuracy. Finally, the distributions of probabilities should be carefully identified since giving the mean and the standard deviation were found to be insufficient.

  4. Carbofuran removal in continuous-photocatalytic reactor: Reactor optimization, rate-constant determination and carbofuran degradation pathway analysis.

    Science.gov (United States)

    Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N

    2017-02-22

    Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO2) catalyst was investigated. The effects of feed flow rate, TiO2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R(2) ∼ 0.964). The addition of 1 mL min(-1) hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L(-1), TiO2 ∼5 mg L(-1) and feed flow rate ∼82.5 mL min(-1). Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.

  5. Seismic fragility analysis of buried steel piping at P, L, and K reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wingo, H.E.

    1989-10-01

    Analysis of seismic strength of buried cooling water piping in reactor areas is necessary to evaluate the risk of reactor operation because seismic events could damage these buried pipes and cause loss of coolant accidents. This report documents analysis of the ability of this piping to withstand the combined effects of the propagation of seismic waves, the possibility that the piping may not behave in a completely ductile fashion, and the distortions caused by relative displacements of structures connected to the piping.

  6. Perturbation analysis of the TRIGA Mark II reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

    2013-04-15

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  7. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  8. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  9. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  10. APPROPRIATE ALLOCATION OF CONTINGENCY USING RISK ANALYSIS METHODOLOGY

    Directory of Open Access Journals (Sweden)

    Andi Andi

    2004-01-01

    Full Text Available Many cost overruns in the world of construction are attributable to either unforeseen events or foreseen events for which uncertainty was not appropriately accommodated. It is argued that a significant improvement to project management performance may result from greater attention to the process of analyzing project risks. The objective of this paper is to propose a risk analysis methodology for appropriate allocation of contingency in project cost estimation. In the first step, project risks will be identified. Influence diagramming technique is employed to identify and to show how the risks affect the project cost elements and also the relationships among the risks themselves. The second step is to assess the project costs with regards to the risks under consideration. Using a linguistic approach, the degree of uncertainty of identified project risks is assessed and quantified. The problem of dependency between risks is taken into consideration during this analysis. For the final step, as the main purpose of this paper, a method for allocating appropriate contingency is presented. Two types of contingencies, i.e. project contingency and management reserve are proposed to accommodate the risks. An illustrative example is presented at the end to show the application of the methodology.

  11. Contemporary Impact Analysis Methodology for Planetary Sample Return Missions

    Science.gov (United States)

    Perino, Scott V.; Bayandor, Javid; Samareh, Jamshid A.; Armand, Sasan C.

    2015-01-01

    Development of an Earth entry vehicle and the methodology created to evaluate the vehicle's impact landing response when returning to Earth is reported. NASA's future Mars Sample Return Mission requires a robust vehicle to return Martian samples back to Earth for analysis. The Earth entry vehicle is a proposed solution to this Mars mission requirement. During Earth reentry, the vehicle slows within the atmosphere and then impacts the ground at its terminal velocity. To protect the Martian samples, a spherical energy absorber called an impact sphere is under development. The impact sphere is composed of hybrid composite and crushable foam elements that endure large plastic deformations during impact and cause a highly nonlinear vehicle response. The developed analysis methodology captures a range of complex structural interactions and much of the failure physics that occurs during impact. Numerical models were created and benchmarked against experimental tests conducted at NASA Langley Research Center. The postimpact structural damage assessment showed close correlation between simulation predictions and experimental results. Acceleration, velocity, displacement, damage modes, and failure mechanisms were all effectively captured. These investigations demonstrate that the Earth entry vehicle has great potential in facilitating future sample return missions.

  12. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  13. Methodology to resolve the transport equation with the discrete ordinates code TORT into the IPEN/MB-01 reactor

    OpenAIRE

    Bernal García, Álvaro; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael

    2014-01-01

    This is an Accepted Manuscript of an article published by Taylor & Francis in International Journal of Computer Mathematics in 2014, available online: http://www.tandfonline.com/10.1080/00207160.2013.799668 Resolution of the steady-state Neutron Transport Equation in a nuclear pool reactor is usually achieved by means of two different numerical methods: Monte Carlo (stochastic) and Discrete Ordinates (deterministic). The Discrete Ordinates method solves the Neutron Transport Equation for a...

  14. Analysis of barium hydroxide and calcium hydroxide slurry carbonation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Patch, K.D.; Hart, R.P.; Schumacher, W.A.

    1980-05-01

    The removal of CO/sub 2/ from air was investigated by using a continuous-agitated-slurry carbonation reactor containing either barium hydroxide (Ba(OH)/sub 2/) or calcium hydroxide (Ca(OH)/sub 2/). Such a process would be applied to scrub /sup 14/CO/sub 2/ from stack gases at nuclear-fuel reprocessing plants. Decontamination factors were characterized for reactor conditions which could alter hydrodynamic behavior. An attempt was made to characterize reactor performance with models assuming both plug flow and various degrees of backmixing in the gas phase. The Ba(OH)/sub 2/ slurry enabled increased conversion, but apparently the process was controlled under some conditions by phenomena differing from those observed for carbonation by Ca(OH)/sub 2/. Overall reaction mechanisms are postulated.

  15. Configurational analysis of an EBT reactor in various magnetic geometries

    Energy Technology Data Exchange (ETDEWEB)

    Owen, L.W.; Uckan, N.A.

    1980-01-01

    Optimization of vacuum field particle confinement in an ELMO Bumpy Torus (EBT) reactor has been considered. Several methods of improving the efficient utilization of magnetic fields and the particle confinement characteristics of a reactor have been analyzed. These include the use of (1) magnets with a large mirror ratio, (2) high field Nb/sub 3/Sn or Nb/sub 3/Sn/NbTi hybrid mirror coils, (3) split-wedge mirror coils, (4) aspect ratio enhancement (ARE) coils, and (5) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbits and three-dimensional tensor pressure equilibrium calculations have shown that the ARE and SYM coils used in conjunction with high field magnets offer the most promise of good plasma performance in a smaller size (up to 50%) EBT reactor. The relative merits of each magnetic configuration are discussed, and the design characteristics are given.

  16. Total loss of AC power analysis for EPR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Darnowski, Piotr, E-mail: piotr.darnowski@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Skrzypek, Eleonora, E-mail: eleonora.skrzypek@ncbj.gov.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); National Centre for Nuclear Research (NCBJ), A. Sołtana 7, 05-400 Otwock-Świerk (Poland); Mazgaj, Piotr, E-mail: piotr.mazgaj@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Świrski, Konrad [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Gandrille, Pascal [AREVA NP SAS, Tour AREVA, 1 place Jean Millier, 92084 Paris La Défense (France)

    2015-08-15

    Highlights: • Total loss of AC power (Station Blackout) was simulated for the EPR reactor model. • In-vessel phase of the accident is under consideration. • Comparison of MELCOR and MAAP results is presented. • MELCOR and MAAP results are comparable. - Abstract: In this paper the results of severe accident simulations for the EPR reactor in the case of loss of offsite power combined with total failure of all diesel generators (total loss of AC power) are presented. Calculations were performed with MELCOR 2.1 computer code for in-vessel phase of the accident. In this scenario, the unavailability of all offsite and onsite power sources and the lack of cooling leads directly to core degradation, material relocation to the lower plenum and rupture of the reactor pressure vessel. MELCOR results were compared qualitatively and quantitatively with MAAP4 code results and show a good agreement.

  17. Plasma engineering analysis of a small torsatron reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lacatski, J.T.; Houlberg, W.A.; Uckan, N.A.

    1985-10-01

    This study examines the plasma physics and reactor engineering feasibility of a small, medium aspect ratio, high-beta, l = 2, D-T torsatron power reactor, based on the magnetic configuration of the Advanced Toroidal Facility, Oak Ridge National Laboratory. Plasma analyses are performed to assess whether confinement in a small, average radius plasma is sufficient to yield an ignited or high-Q driven device. Much of the physics assessment focuses on an evaluation of the radial electric field created by the nonambipolar particle flux. Detailed transport simulations are done with both fixed and self-consistent evolution of the radial electric field. Basic reactor engineering considerations taken into account are neutron wall loading, maximum magnetic field at the helical coils, coil shield thickness, and tritium breeding blanket-shield thickness.

  18. Segment clustering methodology for unsupervised Holter recordings analysis

    Science.gov (United States)

    Rodríguez-Sotelo, Jose Luis; Peluffo-Ordoñez, Diego; Castellanos Dominguez, German

    2015-01-01

    Cardiac arrhythmia analysis on Holter recordings is an important issue in clinical settings, however such issue implicitly involves attending other problems related to the large amount of unlabelled data which means a high computational cost. In this work an unsupervised methodology based in a segment framework is presented, which consists of dividing the raw data into a balanced number of segments in order to identify fiducial points, characterize and cluster the heartbeats in each segment separately. The resulting clusters are merged or split according to an assumed criterion of homogeneity. This framework compensates the high computational cost employed in Holter analysis, being possible its implementation for further real time applications. The performance of the method is measure over the records from the MIT/BIH arrhythmia database and achieves high values of sensibility and specificity, taking advantage of database labels, for a broad kind of heartbeats types recommended by the AAMI.

  19. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  20. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    Energy Technology Data Exchange (ETDEWEB)

    Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

    2016-08-15

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  1. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  2. 2D fluid model analysis for the effect of 3D gas flow on a capacitively coupled plasma deposition reactor

    Science.gov (United States)

    Kim, Ho Jun; Lee, Hae June

    2016-06-01

    The wide applicability of capacitively coupled plasma (CCP) deposition has increased the interest in developing comprehensive numerical models, but CCP imposes a tremendous computational cost when conducting a transient analysis in a three-dimensional (3D) model which reflects the real geometry of reactors. In particular, the detailed flow features of reactive gases induced by 3D geometric effects need to be considered for the precise calculation of radical distribution of reactive species. Thus, an alternative inclusive method for the numerical simulation of CCP deposition is proposed to simulate a two-dimensional (2D) CCP model based on the 3D gas flow results by simulating flow, temperature, and species fields in a 3D space at first without calculating the plasma chemistry. A numerical study of a cylindrical showerhead-electrode CCP reactor was conducted for particular cases of SiH4/NH3/N2/He gas mixture to deposit a hydrogenated silicon nitride (SiN x H y ) film. The proposed methodology produces numerical results for a 300 mm wafer deposition reactor which agree very well with the deposition rate profile measured experimentally along the wafer radius.

  3. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  4. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  5. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  6. Human reliability analysis of the Tehran research reactor using the SPAR-H method

    Directory of Open Access Journals (Sweden)

    Barati Ramin

    2012-01-01

    Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.

  7. Preliminary accident analysis of Flexblue® underwater reactor

    Directory of Open Access Journals (Sweden)

    Haratyk Geoffrey

    2015-01-01

    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  8. Limiting photocurrent analysis of a wide channel photoelectrochemical flow reactor

    Science.gov (United States)

    Davis, Jonathan T.; Esposito, Daniel V.

    2017-03-01

    The development of efficient and scalable photoelectrochemical (PEC) reactors is of great importance for the eventual commercialization of solar fuels technology. In this study, we systematically explore the influence of convective mass transport and light intensity on the performance of a 3D-printed PEC flow cell reactor based on a wide channel, parallel plate geometry. Using this design, the limiting current density generated from the hydrogen evolution reaction at a p-Si metal–insulator–semiconductor (MIS) photocathode was investigated under varied reactant concentration, fluid velocity, and light intensity. Additionally, a simple model is introduced to predict the range of operating conditions (reactant concentration, light intensity, fluid velocity) for which the photocurrent generated in a parallel plate PEC flow cell is limited by light absorption or mass transport. This model can serve as a useful guide for the design and operation of wide-channel PEC flow reactors. The results of this study have important implications for PEC reactors operating in electrolytes with dilute reactant concentrations and/or under high light intensities where high fluid velocities are required in order to avoid operation in the mass transport-limited regime.

  9. Portfolio Assessment on Chemical Reactor Analysis and Process Design Courses

    Science.gov (United States)

    Alha, Katariina

    2004-01-01

    Assessment determines what students regard as important: if a teacher wants to change students' learning, he/she should change the methods of assessment. This article describes the use of portfolio assessment on five courses dealing with chemical reactor and process design during the years 1999-2001. Although the use of portfolio was a new…

  10. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  11. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    Science.gov (United States)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  12. Microbial analysis in biogas reactors suffering by foaming incidents

    DEFF Research Database (Denmark)

    Kougias, Panagiotis; De Francisci, Davide; Treu, Laura

    2014-01-01

    Foam formation can lead to total failure of digestion process in biogas plants. In the present study, possible correlation between foaming and the presence of specific microorganisms in biogas reactors was elucidated. The microbial ecology of continuous fed digesters overloaded with proteins...

  13. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  14. Temperature Fluctuation Characteristics Analysis for Steam Generator of Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    ZHU; Li-na; WU; Zhi-guang

    2015-01-01

    In the case of boiling heat transfer deterioration,temperature fluctuating may accelerate the corrosion of heat transfer tubes and can also lead to thermal stress on the tubes.In this paper,dryout-induced temperature fluctuation for the fast reactor steam generator is investigated.The impacts of water flow rate,sodium inlet temperature and the outlet steam

  15. Urban transportation energy conservation: case city applications of analysis methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Atherton, T.J.; Suhrbier, J.H.

    1979-10-01

    This report demonstrates the application of a set of disaggregate travel demand methodologies to the analysis of potential energy conservation strategies in three urban areas: Denver, Colorado; Fort Worth, Texas; and San Francisco, California. The methodologies are sketch planning in nature and include the forecasting of changes in automobile ownership; work trip model shares by drive alone, shared ride, and transit; non-work trip frequency, destination, and mode choice; fuel consumption and vehicle emissions. Using concepts of market segmentation and random sample household aggregation, both computer system and manual worksheet versions of the basic approach are applied. Policies analyzed include those related as employer based ride sharing, parking management, transit, pricing, and traffic operations. Considerable variations in the potential effectiveness was found, among the three urban areas, depending in large part on the availability of alternative travel modes such as transit. Descriptions are provided of the individual policy analyses performed, the methods by which example policies were analyzed, the necessary data preparation activities, and the procedures used to adapt the set of travel demand models to the unique conditions of each of the three metropolitan areas.

  16. Analysis of selection methodologies for combinatorial library design.

    Science.gov (United States)

    Pascual, Rosalia; Borrell, José I; Teixidó, Jordi

    2003-01-01

    We have implemented and adapted in Pralins (Program for Rational Analysis of Libraries in silico), the most popular sparse (cherry picking) and full array (sublibrary) selection algorithms: hierarchical clustering, k-means clustering, Optimum Binning, Jarvis Patrick, Pral-SE (partitioning techniques) and MaxSum, MaxMin, MaxMin averaged, DN2, CTD (distance-based methods). We have validated the program with an already synthesized three-component combinatorial library of FXR partial agonists characterized by standard computational chemistry descriptors as case study. This has let us analyze the goodness of both the partitioning techniques for space division and all the selection methodologies with respect to representativity in terms of population and space coverage for different selection sizes. Within the chemical space analyzed, both hierarchical clustering and Optimum Binning division strategies are found to be the most advantageous reference space divisions to be used in the subsequent population and space coverage studies. Complete hierarchical clustering appears also to be the preferred selection methodology for both sparse and full array problems. The full array restriction fulfillment can easily be overcome by convenient optimization algorithms that allow optimal reagent selection preserving > 90% of the population coverage.

  17. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  18. A New Methodology of Spatial Cross-Correlation Analysis

    Science.gov (United States)

    Chen, Yanguang

    2015-01-01

    Spatial correlation modeling comprises both spatial autocorrelation and spatial cross-correlation processes. The spatial autocorrelation theory has been well-developed. It is necessary to advance the method of spatial cross-correlation analysis to supplement the autocorrelation analysis. This paper presents a set of models and analytical procedures for spatial cross-correlation analysis. By analogy with Moran’s index newly expressed in a spatial quadratic form, a theoretical framework is derived for geographical cross-correlation modeling. First, two sets of spatial cross-correlation coefficients are defined, including a global spatial cross-correlation coefficient and local spatial cross-correlation coefficients. Second, a pair of scatterplots of spatial cross-correlation is proposed, and the plots can be used to visually reveal the causality behind spatial systems. Based on the global cross-correlation coefficient, Pearson’s correlation coefficient can be decomposed into two parts: direct correlation (partial correlation) and indirect correlation (spatial cross-correlation). As an example, the methodology is applied to the relationships between China’s urbanization and economic development to illustrate how to model spatial cross-correlation phenomena. This study is an introduction to developing the theory of spatial cross-correlation, and future geographical spatial analysis might benefit from these models and indexes. PMID:25993120

  19. Multi-Reactor Transmutation Analysis Utility (MRTAU,alpha1): Verification

    Energy Technology Data Exchange (ETDEWEB)

    Andrea Alfonsi; Samuel E. Bays; Cristian Rabiti; Steven J. Piet

    2011-02-01

    Multi-Reactor Transmutation Utility (MRTAU) is a general depletion/decay algorithm under development at INL to support quick assessment of off-normal fuel cycle scenarios of similar nature to well studied reactor and fuel cycle concepts for which isotopic and cross-section data exists. MRTAU has been used in the past for scoping calculations to determine actinide composition evolution over the course of multiple recycles in Light Water Reactor Mixed Oxide and Sodium cooled Fast Reactor. In these applications, various actinide partitioning scenarios of interest were considered. The code has recently been expanded to include fission product generation, depletion and isotopic evolution over multiple recycles. The capability was added to investigate potential partial separations and/or limited recycling technologies such as Melt-Refining, AIROX, DUPIC or other fuel recycle technology where the recycled fuel stream is not completely decontaminated of fission products prior to being re-irradiated in a subsequent reactor pass. This report documents the code's solution methodology and algorithm as well as its solution accuracy compared to the SCALE6.0 software suite.

  20. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  1. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  2. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  3. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  4. Methodological developments in the field of structural integrity analyses of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available Buildings, structures and systems of large scale and high value (e.g. conventional and nuclear power plants, etc. are designed for a certain, limited service lifetime. If the standards and guidelines of the time are taken into account during the design process, the resulting structures will operate safely in most cases. However, in the course of technical history there were examples of unusual, catastrophic failures of structures, even resulting in human casualties. Although the concept of Structural Integrity first appeared in industrial applications only two-three decades ago, its pertinence has been growing higher ever since. Four nuclear power generation units have been constructed in Hungary, more than 30 years ago. In every unit, VVER-440 V213 type light-water cooled, light-water moderated, pressurized water reactors are in operation. Since the mid-1980s, Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPV have been conducted in Hungary, where the concept of structural integrity was the basis of research and development. In the first part of the paper, a short historic overview is given, where the origins of the Structural Integrity concept are presented, and the beginnings of Structural Integrity in Hungary are summarized. In the second part, a new conceptual model of Structural Integrity is introduced. In the third part, a brief description of the VVER-440 V213 type RPV and its surrounding primary system is presented. In the fourth part, a conceptual model developed for PTS Structural Integrity Analyses is explained.

  5. Methodology for Modeling and Analysis of Business Processes (MMABP

    Directory of Open Access Journals (Sweden)

    Vaclav Repa

    2015-10-01

    Full Text Available This paper introduces the methodology for modeling business processes. Creation of the methodology is described in terms of the Design Science Method. Firstly, the gap in contemporary Business Process Modeling approaches is identified and general modeling principles which can fill the gap are discussed. The way which these principles have been implemented in the main features of created methodology is described. Most critical identified points of the business process modeling are process states, process hierarchy and the granularity of process description. The methodology has been evaluated by use in the real project. Using the examples from this project the main methodology features are explained together with the significant problems which have been met during the project. Concluding from these problems together with the results of the methodology evaluation the needed future development of the methodology is outlined.

  6. 3D TRANSIENT COUPLED THERMO-ELASTIC-PLASTIC CONTACT SEALING ANALYSIS OF REACTOR PRESSURE VESSEL

    Institute of Scientific and Technical Information of China (English)

    Du Xuesong; Li Runfang; Lin Tengjiao

    2005-01-01

    Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.

  7. Methodology for the analysis and retirement of assets: Power transformers

    Directory of Open Access Journals (Sweden)

    Gustavo Adolfo Gómez-Ramírez

    2015-09-01

    Full Text Available The following article consists of the development of a methodology of the area of the engineering of high voltage for the analysis and retirement of repaired power transformers, based on engineering criteria, in order to establish a correlation between the conditions of the transformer from several points of view: electrical, mechanical, dielectrics and thermal. Realizing an analysis of the condition of the question, there are two situations of great transcendency, first one, the international procedure are a “guide” for the acceptance of new transformers, by what they cannot be applied to the letter for repaired transformers, due to the process itself of degradation that the transformer has suffered with to happen of the years and all the factors that they were carrying to a possible repair. Second one, with base in the most recent technical literature, there have been analyzed articles, which analyze the oil dielectrics and the paper, which correlations are established between the quality of the insulating paper and the furan concentrations in the oils. To finish, great part of the investigation till now realized, it has focused in the analysis of the transformer, from the condition of the dielectric oil, so in most cases, there is not had the possibility of realizing a forensic engineering inside the transformer in operation and of being able this way, of analyzing the components of design who can compromise the integrity and operability of the same one.

  8. Activation analysis with the Michigan Reactor: application to Michigan problems

    Energy Technology Data Exchange (ETDEWEB)

    Nass, H. W.

    1963-10-15

    A review is presented on activation analyses performed by the Michigan reactor. Typical determinations include: sodium deposited on gold leaf for nuclear physics experiments; cobalt and silver in high purity copper used in electron diffraction studies; sodium, manganese and chlorine in experimental reactor grade terphenyl; zirconium in an experimental tungsten - chromium alloy; argon, vanadium, manganese, chlorine, scandium, and iron in spectroscopic grade carbon rod; vanadium and aluminum in synthetically grown sapphires used for x-ray fluorescence studies; hafnium in uranyl nitrate; aluminum in silicon catalyst; vanadium in synthetic cadmium--telluride crystals used in electron spin resonance studies; and sodium and potassium in human tissue from internal organs, Analyses on oil samples and cadmium rod have also been performed. (P.C.H.)

  9. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  10. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  11. Microbial analysis in biogas reactors suffering by foaming incidents.

    Science.gov (United States)

    Kougias, Panagiotis G; De Francisci, Davide; Treu, Laura; Campanaro, Stefano; Angelidaki, Irini

    2014-09-01

    Foam formation can lead to total failure of digestion process in biogas plants. In the present study, possible correlation between foaming and the presence of specific microorganisms in biogas reactors was elucidated. The microbial ecology of continuous fed digesters overloaded with proteins, lipids and carbohydrates before and after foaming incidents was characterized using 16S rRNA gene sequencing. Moreover, the microbial diversity between the liquid and foaming layer was assessed. A number of genera that are known to produce biosurfactants, contain mycolic acid in their cell wall, or decrease the surface tension of the media, increased their relative abundance after foam formation. Finally, a microorganism similar to widely known foaming bacteria (Nocardia and Desulfotomaculum) was found to increase its relative abundance in all reactors once foam was observed, regardless of the used substrate. These findings suggest that foaming and specific microorganisms might have direct association which requires to be further investigated.

  12. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  13. Response Surface Methodology Control Rod Position Optimization of a Pressurized Water Reactor Core Considering Both High Safety and Low Energy Dissipation

    Directory of Open Access Journals (Sweden)

    Yi-Ning Zhang

    2017-02-01

    Full Text Available Response Surface Methodology (RSM is introduced to optimize the control rod positions in a pressurized water reactor (PWR core. The widely used 3D-IAEA benchmark problem is selected as the typical PWR core and the neutron flux field is solved. Besides, some additional thermal parameters are assumed to obtain the temperature distribution. Then the total and local entropy production is calculated to evaluate the energy dissipation. Using RSM, three directions of optimization are taken, which aim to determine the minimum of power peak factor Pmax, peak temperature Tmax and total entropy production Stot. These parameters reflect the safety and energy dissipation in the core. Finally, an optimization scheme was obtained, which reduced Pmax, Tmax and Stot by 23%, 8.7% and 16%, respectively. The optimization results are satisfactory.

  14. A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2011-07-01

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

  15. HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel

    Energy Technology Data Exchange (ETDEWEB)

    Noguera Oliva, O.

    2011-07-01

    The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.

  16. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  17. Thermal analysis of IRT-T reactor fuel elements

    OpenAIRE

    Naymushin, Artem Georgievich; Chertkov, Yuri Borisovich; Lebedev, Ivan Igorevich; Anikin, Mikhail Nikolaevich

    2015-01-01

    The article describes the method and results of thermo-physical calculations of IRT-T reactor core. Heat fluxes, temperatures of cladding, fuel meat and coolant were calculated for height of core, azimuth directions of FA and each fuel elements in FA. Average calculated values of uniformity factor of energy release distribution for height of fuel assemblies were shown in this research. Onset nucleate boiling temperature and ONB-ratio were calculated. Shows that temperature regimes of fuel ele...

  18. Microbial Community Analysis of Anaerobic Reactors Treating Soft Drink Wastewater

    Science.gov (United States)

    Narihiro, Takashi; Kim, Na-Kyung; Mei, Ran; Nobu, Masaru K.; Liu, Wen-Tso

    2015-01-01

    The anaerobic packed-bed (AP) and hybrid packed-bed (HP) reactors containing methanogenic microbial consortia were applied to treat synthetic soft drink wastewater, which contains polyethylene glycol (PEG) and fructose as the primary constituents. The AP and HP reactors achieved high COD removal efficiency (>95%) after 80 and 33 days of the operation, respectively, and operated stably over 2 years. 16S rRNA gene pyrotag analyses on a total of 25 biofilm samples generated 98,057 reads, which were clustered into 2,882 operational taxonomic units (OTUs). Both AP and HP communities were predominated by Bacteroidetes, Chloroflexi, Firmicutes, and candidate phylum KSB3 that may degrade organic compound in wastewater treatment processes. Other OTUs related to uncharacterized Geobacter and Spirochaetes clades and candidate phylum GN04 were also detected at high abundance; however, their relationship to wastewater treatment has remained unclear. In particular, KSB3, GN04, Bacteroidetes, and Chloroflexi are consistently associated with the organic loading rate (OLR) increase to 1.5 g COD/L-d. Interestingly, KSB3 and GN04 dramatically decrease in both reactors after further OLR increase to 2.0 g COD/L-d. These results indicate that OLR strongly influences microbial community composition. This suggests that specific uncultivated taxa may take central roles in COD removal from soft drink wastewater depending on OLR. PMID:25748027

  19. Microbial community analysis of anaerobic reactors treating soft drink wastewater.

    Directory of Open Access Journals (Sweden)

    Takashi Narihiro

    Full Text Available The anaerobic packed-bed (AP and hybrid packed-bed (HP reactors containing methanogenic microbial consortia were applied to treat synthetic soft drink wastewater, which contains polyethylene glycol (PEG and fructose as the primary constituents. The AP and HP reactors achieved high COD removal efficiency (>95% after 80 and 33 days of the operation, respectively, and operated stably over 2 years. 16S rRNA gene pyrotag analyses on a total of 25 biofilm samples generated 98,057 reads, which were clustered into 2,882 operational taxonomic units (OTUs. Both AP and HP communities were predominated by Bacteroidetes, Chloroflexi, Firmicutes, and candidate phylum KSB3 that may degrade organic compound in wastewater treatment processes. Other OTUs related to uncharacterized Geobacter and Spirochaetes clades and candidate phylum GN04 were also detected at high abundance; however, their relationship to wastewater treatment has remained unclear. In particular, KSB3, GN04, Bacteroidetes, and Chloroflexi are consistently associated with the organic loading rate (OLR increase to 1.5 g COD/L-d. Interestingly, KSB3 and GN04 dramatically decrease in both reactors after further OLR increase to 2.0 g COD/L-d. These results indicate that OLR strongly influences microbial community composition. This suggests that specific uncultivated taxa may take central roles in COD removal from soft drink wastewater depending on OLR.

  20. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  1. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    Science.gov (United States)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  2. Supporting Space Systems Design via Systems Dependency Analysis Methodology

    Science.gov (United States)

    Guariniello, Cesare

    assess the behavior of each system based on its internal status and on the topology of its dependencies on systems connected to it. Designers and decision makers can therefore quickly analyze and explore the behavior of complex systems and evaluate different architectures under various working conditions. The methods support educated decision making both in the design and in the update process of systems architecture, reducing the need to execute extensive simulations. In particular, in the phase of concept generation and selection, the information given by the methods can be used to identify promising architectures to be further tested and improved, while discarding architectures that do not show the required level of global features. The methods, when used in conjunction with appropriate metrics, also allow for improved reliability and risk analysis, as well as for automatic scheduling and re-scheduling based on the features of the dependencies and on the accepted level of risk. This dissertation illustrates the use of the two methods in sample aerospace applications, both in the operational and in the developmental domain. The applications show how to use the developed methodology to evaluate the impact of failures, assess the criticality of systems, quantify metrics of interest, quantify the impact of delays, support informed decision making when scheduling the development of systems and evaluate the achievement of partial capabilities. A larger, well-framed case study illustrates how the Systems Operational Dependency Analysis method and the Systems Developmental Dependency Analysis method can support analysis and decision making, at the mid and high level, in the design process of architectures for the exploration of Mars. The case study also shows how the methods do not replace the classical systems engineering methodologies, but support and improve them.

  3. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  4. 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.

  5. 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.

  6. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  7. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    Energy Technology Data Exchange (ETDEWEB)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    2001-07-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  8. Rate-Only analysis with reactor-off data in the Double Chooz experiment

    CERN Document Server

    Novella, P

    2013-01-01

    Among ongoing reactor-based experiments, Double Chooz is unique in obtaining data when the reactor cores are brought down for maintenance. These reactor-off data allow for a clean measurement of the backgrounds of the experiment, thus being of uppermost importance for the theta13 oscillation analysis. While the oscillation results published by the collaboration in 2011 and 2012 rely on background models derived from reactor-on data, in this talk we present an independent study based on the handle provided by 7.53 days of reactor-off data. A global fit to both theta13 and the total background is performed by analyzing the observed neutrino rate as a function of the non-oscillated expected rate for different reactor power conditions. The result presented in this talk is fully consistent with the one already published by Double Chooz. As they both yield almost the same precision, this work stands as a prove of the reliability of the background estimates and the oscillation analysis of the experiment.

  9. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  10. Methodology for Modeling and Analysis of Business Processes (MMABP)

    OpenAIRE

    Vaclav Repa; Tomas Bruckner

    2015-01-01

    This paper introduces the methodology for modeling business processes. Creation of the methodology is described in terms of the Design Science Method. Firstly, the gap in contemporary Business Process Modeling approaches is identified and general modeling principles which can fill the gap are discussed. The way which these principles have been implemented in the main features of created methodology is described. Most critical identified points of the business process modeling are process stat...

  11. FRACTAL ANALYSIS OF TRABECULAR BONE: A STANDARDISED METHODOLOGY

    Directory of Open Access Journals (Sweden)

    Ian Parkinson

    2011-05-01

    Full Text Available A standardised methodology for the fractal analysis of histological sections of trabecular bone has been established. A modified box counting method has been developed for use on a PC based image analyser (Quantimet 500MC, Leica Cambridge. The effect of image analyser settings, magnification, image orientation and threshold levels, was determined. Also, the range of scale over which trabecular bone is effectively fractal was determined and a method formulated to objectively calculate more than one fractal dimension from the modified Richardson plot. The results show that magnification, image orientation and threshold settings have little effect on the estimate of fractal dimension. Trabecular bone has a lower limit below which it is not fractal (λ<25 μm and the upper limit is 4250 μm. There are three distinct fractal dimensions for trabecular bone (sectional fractals, with magnitudes greater than 1.0 and less than 2.0. It has been shown that trabecular bone is effectively fractal over a defined range of scale. Also, within this range, there is more than 1 fractal dimension, describing spatial structural entities. Fractal analysis is a model independent method for describing a complex multifaceted structure, which can be adapted for the study of other biological systems. This may be at the cell, tissue or organ level and compliments conventional histomorphometric and stereological techniques.

  12. Methodological frontier in operational analysis for roundabouts: a review

    Directory of Open Access Journals (Sweden)

    Orazio Giuffre'

    2016-11-01

    Full Text Available Several studies and researches have shown that modern roundabouts are safe and effective as engineering countermeasures for traffic calming, and they are now widely used worldwide. The increasing use of roundabouts and, more recently, turbo and flower roundabouts, has induced a great variety of experiences in the field of intersection design, traffic safety and capacity modelling. As for unsignalized intersections which represent the starting point to extend knowledge about the operational analysis to roundabouts, the general situation in capacity estimation is still characterized by the discussion between gap acceptance models and empirical regression models. However, capacity modelling must contain both the analytical construction and then solution of the model, and the implementation of driver behavior. Thus, issues on a realistic modelling of driver behavior by the parameters that are included into the models are always of interest for practioners and analysts in transportation and road infrastructure engineering. Based on these considerations, this paper presents a literature review about the key methodological issues in the operational analysis of modern roundabouts. Focus is made on the aspects associated with the gap acceptance behavior, the derivation of the analytical-based models and the calculation of parameters included into the capacity equations, as well as steady state and non-steady state conditions and uncertainty in entry capacity estimation. At last, insights on future developments of the research in this field of investigation will be also outlined.

  13. Development and analysis of some versions of the fractional-order point reactor kinetics model for a nuclear reactor with slab geometry

    Science.gov (United States)

    Vyawahare, Vishwesh A.; Nataraj, P. S. V.

    2013-07-01

    In this paper, we report the development and analysis of some novel versions and approximations of the fractional-order (FO) point reactor kinetics model for a nuclear reactor with slab geometry. A systematic development of the FO Inhour equation, Inverse FO point reactor kinetics model, and fractional-order versions of the constant delayed neutron rate approximation model and prompt jump approximation model is presented for the first time (for both one delayed group and six delayed groups). These models evolve from the FO point reactor kinetics model, which has been derived from the FO Neutron Telegraph Equation for the neutron transport considering the subdiffusive neutron transport. Various observations and the analysis results are reported and the corresponding justifications are addressed using the subdiffusive framework for the neutron transport. The FO Inhour equation is found out to be a pseudo-polynomial with its degree depending on the order of the fractional derivative in the FO model. The inverse FO point reactor kinetics model is derived and used to find the reactivity variation required to achieve exponential and sinusoidal power variation in the core. The situation of sudden insertion of negative reactivity is analyzed using the FO constant delayed neutron rate approximation. Use of FO model for representing the prompt jump in reactor power is advocated on the basis of subdiffusion. Comparison with the respective integer-order models is carried out for the practical data. Also, it has been shown analytically that integer-order models are a special case of FO models when the order of time-derivative is one. Development of these FO models plays a crucial role in reactor theory and operation as it is the first step towards achieving the FO control-oriented model for a nuclear reactor. The results presented here form an important step in the efforts to establish a step-by-step and systematic theory for the FO modeling of a nuclear reactor.

  14. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  15. Development of guidelines for inelastic analysis in design of fast reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Kyotada [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Kasahara, Naoto [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan)], E-mail: kasahara.naoto@jaea.go.jp; Morishita, Masaki [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Shibamoto, Hiroshi; Inoue, Kazuhiko [The Japan Atomic Power Company, Dispatched to JAEA, 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Nakayama, Yasunari [Kawasaki Heavy Industries, Ltd., 1-1, Kawasaki, Akashi, Hyogo 673-8666 (Japan)

    2008-02-15

    The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users.

  16. Modern design and safety analysis of the University of Florida Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kjordan@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Springfels, D., E-mail: dspringfels@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Schubring, D., E-mail: dlschubring@ufl.edu [University of Florida, 202 Nuclear Science Building, PO Box 118300, Gainesville, FL 32611-8300 (United States)

    2015-05-15

    Highlights: • A new safety analysis of the University of Florida Training Reactor is presented. • This analysis uses modern codes and replaces the NRC approved analysis from 1982. • Reduction in engineering margin confirms that the UFTR is a negligible risk reactor. • Safety systems are not required to ensure that safety limits are not breached. • Negligible risk reactors are ideal for testing digital I&C equipment. - Abstract: A comprehensive series of neutronics and thermal hydraulics analyses were conducted to demonstrate the University of Florida Training Reactor (UFTR), an ARGONAUT type research reactor, as a negligible risk reactor that does not require safety-related systems or components to prevent breach of a safety limit. These analyses show that there is no credible UFTR accident that would result in major fuel damage or risk to public health and safety. The analysis was based on two limiting scenarios, whose extremity bound all other accidents of consequence: (1) the large step insertion of positive reactivity and (2) the release of fission products due to mechanical damage to a spent fuel plate. The maximum step insertion of positive reactivity was modeled using PARET/ANL software and shows a maximum peak fuel temperature of 283.2 °C, which is significantly below the failure limit of 530 °C. The exposure to the staff and general public was calculated for the worst-case fission product release scenario using the ORIGEN-S and COMPLY codes and was shown to be 6.5% of the annual limit. Impacts on reactor operations and an Instrumentation & Control System (I&C) upgrade are discussed.

  17. A study on safety analysis methodology in spent fuel dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

    2004-02-15

    Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

  18. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  19. Steady thermal hydraulic analysis for a molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    ZHANG Dalin; QIU Suizheng; LIU Changliang; SU Guanghui

    2008-01-01

    The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.

  20. 50 Years of coastal erosion analysis: A new methodological approach.

    Science.gov (United States)

    Prieto Campos, Antonio; Diaz Cuevas, Pilar; Ojeda zujar, Jose; Guisado-Pintado, Emilia

    2017-04-01

    Coasts over the world have been subjected to increased anthropogenic pressures which combined with natural hazards impacts (storm events, rising sea-levels) have led to strong erosion problems with negative impacts on the economy and the safety of coastal communities. The Andalusian coast (South Spain) is a renowned global tourist destination. In the past decades a deep transformation in the economic model led to significant land use changes: strong regulation of rivers, urbanisation and occupation of dunes, among others. As a result irreversible transformations on the coastline, from the aggressive urbanisation undertaken, are now to be faced by local authorities and suffered by locals and visitors. Moreover, the expected impacts derived from the climate change aggravated by anthropic activities emphasises the need for tools that facilitates decision making for a sustainable coastal management. In this contribution a homogeneous (only a proxy and one photointerpreter) methodology is proposed for the calculation of coastal erosion rates of exposed beaches in Andalusia (640 km) through the use of detailed series (1:2500) of open source orthophotographies for the period (1956-1977-2001-2011). The outstanding combination of the traditional software DSAS (Digital Shoreline Analysis System) with a spatial database (PostgreSQL) which integrates the resulting erosion rates with related coastal thematic information (geomorphology, presence of engineering infrastructures, dunes and ecosystems) enhances the capacity of analysis and exploitation. Further, the homogeneity of the method used allows the comparison of the results among years in a highly diverse coast, with both Mediterranean and Atlantic façades. The novelty development and integration of a PostgreSQL/Postgis database facilitates the exploitation of the results by the user (for instance by relating calculated rates with other thematic information as geomorphology of the coast or the presence of a dune field on

  1. Analysis of gallium arsenide deposition in a horizontal chemical vapor deposition reactor using massively parallel computations

    Energy Technology Data Exchange (ETDEWEB)

    Salinger, A.G.; Shadid, J.N.; Hutchinson, S.A. [and others

    1998-01-01

    A numerical analysis of the deposition of gallium from trimethylgallium (TMG) and arsine in a horizontal CVD reactor with tilted susceptor and a three inch diameter rotating substrate is performed. The three-dimensional model includes complete coupling between fluid mechanics, heat transfer, and species transport, and is solved using an unstructured finite element discretization on a massively parallel computer. The effects of three operating parameters (the disk rotation rate, inlet TMG fraction, and inlet velocity) and two design parameters (the tilt angle of the reactor base and the reactor width) on the growth rate and uniformity are presented. The nonlinear dependence of the growth rate uniformity on the key operating parameters is discussed in detail. Efficient and robust algorithms for massively parallel reacting flow simulations, as incorporated into our analysis code MPSalsa, make detailed analysis of this complicated system feasible.

  2. Application of neutron activation analysis system in Xi'an pulsed reactor

    CERN Document Server

    Zhang Wen Shou; Yu Qi

    2002-01-01

    Neutron Activation Analysis System in Xi'an Pulsed Reactor is consist of rabbit fast radiation system and experiment measurement system. The functions of neutron activation analysis are introduced. Based on the radiation system. A set of automatic data handling and experiment simulating system are built. The reliability of data handling and experiment simulating system had been verified by experiment

  3. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  4. Development of trip coverage analysis methodology - CATHENA trip coverage analysis model

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Ho; Ohn, M. Y.; Cho, C. H.; Huh, J. Y.; Na, Y. H.; Lee, S. Y.; Kim, B. G.; Kim, H. H.; Kim, S. W.; Bae, C. J.; Kim, T. M.; Kim, S. R.; Han, B. S.; Moon, B. J.; Oh, M. T. [Korea Power Engineering Co., Yongin (Korea)

    2001-05-01

    This report describes the CATHENA model for trip coverage analysis. This model is prepared based on the Wolsong 2 design data and consist of primary heat transport system, shutdown system, steam and feedwater system, reactor regulating system, heat transport pressure and inventory control system, and steam generator level and pressure control system. The new features and modified parts from the Wolsong 2 CATHENA LOCA Model required for trip coverage analysis is described. this model is tested by simulation of steady state at 100 % FP and at several low powers. Also, the cases of power rundown and power runup are tested. 17 refs., 124 figs., 19 tabs. (Author)

  5. Hazardous materials transportation: a risk-analysis-based routing methodology.

    Science.gov (United States)

    Leonelli, P; Bonvicini, S; Spadoni, G

    2000-01-07

    This paper introduces a new methodology based on risk analysis for the selection of the best route for the transport of a hazardous substance. In order to perform this optimisation, the network is considered as a graph composed by nodes and arcs; each arc is assigned a cost per unit vehicle travelling on it and a vehicle capacity. After short discussion about risk measures suitable for linear risk sources, the arc capacities are introduced by comparison between the societal and individual risk measures of each arc with hazardous materials transportation risk criteria; then arc costs are defined in order to take into account both transportation out-of-pocket expenses and risk-related costs. The optimisation problem can thus be formulated as a 'minimum cost flow problem', which consists of determining for a specific hazardous substance the cheapest flow distribution, honouring the arc capacities, from the origin nodes to the destination nodes. The main features of the optimisation procedure, implemented on the computer code OPTIPATH, are presented. Test results about shipments of ammonia are discussed and finally further research developments are proposed.

  6. RESPONSE SURFACE METHODOLOGY ANALYSIS OF POLYPHENOL RECOVERY FROM ARTICHOKE WASTE

    Directory of Open Access Journals (Sweden)

    Antonio Zuorro

    2014-01-01

    Full Text Available Large amounts of a solid waste consisting mainly of outer bracts and stems are produced from the industrial processing of artichokes. In this study, the recovery of polyphenols from the two waste components was investigated. Extraction experiments were carried outby an environmentally friendly procedure using aqueous ethanol as solvent. The total polyphenol content, expressed as mg of GAE per g of dry weight, was 10.23±0.68 mg/g for bracts and 16.36±0.85 mg/g for stems. To evaluate the effect of Temperature (T, Extraction time (E and liquid-to-solid Ratio (R on the extraction yields, a central composite design coupled with response surface methodology was used. Under the best conditions (T = 50°C, E = 110.4 min and R = 20 mL g-1, extraction yields between 90 and 93% were obtained. Statistical analysis of the data showed that E was the most influential factor, followed by T and R. Simplified polynomial models were developed to describe the effect of individual factors and their interactions on the extraction yield of polyphenols. Overall, the results of this study support the potential of using artichoke waste as a source of natural phenolic antioxidants and give useful directions on how to improve recovery by proper selection of extraction conditions.

  7. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)

    2016-06-15

    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  8. Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors

    Science.gov (United States)

    Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.

    2017-01-01

    Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..

  9. Using functional analysis in archival appraisal a practical and effective alternative to traditional appraisal methodologies

    CERN Document Server

    Robyns, Marcus C

    2014-01-01

    In an age of scarcity and the challenge of electronic records, can archivists and records managers continue to rely upon traditional methodology essentially unchanged since the early 1950s? Using Functional Analysis in Archival Appraisal: A Practical and Effective Alternative to Traditional Appraisal Methodologies shows how archivists in other countries are already using functional analysis, which offers a better, more effective, and imminently more practical alternative to traditional appraisal methodologies that rely upon an analysis of the records themselves.

  10. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  11. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.

  12. Radiochemical analysis of concrete samples for decommission of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zapata-Garcia, Daniel; Wershofen, Herbert [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100 38116, Braunschweig (Germany); Larijani, Cyrus; Sobrino-Petrirena, Maitane; Garcia-Miranda, Maria; Jerome, Simon M. [National Physical Laboratory (NPL), Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom)

    2014-07-01

    Decommissioning of the oldest nuclear power reactors are some of the most challenging technological legacy issues many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Approximately 10 % of the contaminated material ends up in different repositories (depending on their level of contamination) while the rest is decontaminated, measured and released into the environment or sent for recycling. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. The work we present deals with the development and the validation of a procedure for the simultaneous determination of {sup 241}Am, plutonium isotopes, uranium isotopes and {sup 90}Sr in concrete samples. Samples are firstly ground and fused with LiBO{sub 2} and Li{sub 2}B{sub 4}O{sub 7}. After dissolution of the fused sample, silicate and alkaline elements are removed followed by radiochemical separation of the target radionuclides using extraction chromatography. Measurement

  13. Microgenetic Analysis of Moral Development: Theoretical and Methodological Issues

    OpenAIRE

    Barrios, Alia; Barbato,Silviane; Branco,Angela

    2012-01-01

    New ideas and methodologies need to be developed to advance our knowledge in the understanding of moral development. The intertwined nature of human activities, communication processes, and the numerous aspects of morality pose a challenge to researchers to construct a methodology that takes into account cognition, affect, sociocultural processes and characteristics, as well as the active role of individuals in their own development. In this paper we aim at suggesting fresh theoretical ideas ...

  14. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    Science.gov (United States)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  15. Cellulose Isolation Methodology for NMR Analysis of Cellulose Ultrastructure

    Directory of Open Access Journals (Sweden)

    Art J. Ragauskas

    2011-11-01

    Full Text Available In order to obtain accurate information about the ultrastructure of cellulose from native biomass by 13C cross polarization magic angle spinning (CP/MAS NMR spectroscopy the cellulose component must be isolated due to overlapping resonances from both lignin and hemicellulose. Typically, cellulose isolation has been achieved via holocellulose pulping to remove lignin followed by an acid hydrolysis procedure to remove the hemicellulose components. Using 13C CP/MAS NMR and non-linear line-fitting of the cellulose C4 region, it was observed that the standard acid hydrolysis procedure caused an apparent increase in crystallinity of ~10% or less on the cellulose isolated from Populus holocellulose. We have examined the effect of the cellulose isolation method, particularly the acid treatment time for hemicellulose removal, on cellulose ultrastructural characteristics by studying these effects on cotton, microcrystalline cellulose (MCC and holocellulose pulped Populus. 13C CP/MAS NMR of MCC indicated that holocellulose pulping and acid hydrolysis has little effect on the crystalline ultrastructural components of cellulose. Although any chemical method to isolate cellulose from native biomass will invariably alter substrate characteristics, especially those related to regions accessible to solvents, we found those changes to be minimal and consistent in samples of typical crystallinity and lignin/hemicellulose content. Based on the rate of the hemicellulose removal, as determined by HPLC-carbohydrate analysis and magnitude of cellulose ultrastructural alteration, the most suitable cellulose isolation methodology utilizes a treatment of 2.5 M HCl at 100 °C for a standard residence time between 1.5 and 4 h. However, for the most accurate crystallinity results this residence time should be determined empirically for a particular sample.

  16. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  17. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  18. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  19. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  20. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  1. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  2. Energy analysis for the production of biodiesel in a spiral reactor using supercritical tert-butyl methyl ether (MTBE).

    Science.gov (United States)

    Farobie, Obie; Matsumura, Yukihiko

    2015-11-01

    In this study, energy analysis was conducted for the production of biodiesel in a spiral reactor using supercritical tert-butyl methyl ether (MTBE). This study aims to determine the net energy ratio (NER) and energy efficiency for the production of biodiesel using supercritical MTBE and to verify the effectiveness of the spiral reactor in terms of heat recovery efficiency. The analysis results revealed that the NER for this process was 0.92. Meanwhile, the energy efficiency was 0.98, indicating that the production of biodiesel in a spiral reactor using supercritical MTBE is an energy-efficient process. By comparing the energy supply required for biodiesel production between spiral and conventional reactors, the spiral reactor was more efficient than the conventional reactor.

  3. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily

  4. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  5. Review of methodology for accident consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Strenge, D.L.; Soldat, J.K.; Watson, E.C.

    1978-09-01

    This report reviews current methodologies for reactor accident consequence analysis and describes areas where modifications are warranted. Methodologies reviewed are: (1) Models in Regulatory Guides 1.109, 1.111 and 1.113 used for evaluation of compliance with 10 CFR 50 Appendix I; (2) Models in Regulatory Guides used for evaluation of consequences from accidents of Classes 3-8; (3) Models for evaluation of Class 9 accidents presented in the Reactor Safety Study; and (4) Models in the Liquid Pathway Generic Study. The review is designed to aid in the ultimate goal of selection of a comprehensive set of models to extend the Class 9 methodology of the Reactor Safety Study to the analysis of Classes 3-8 accidents.

  6. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-03-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

  7. CFD analysis and flow model reduction for surfactant production in helix reactor

    NARCIS (Netherlands)

    Nikačević, N.M.; Thielen, L.; Twerda, A.; Hof, P.M.J. van den

    2014-01-01

    Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR) were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD) simulations. Non-reactive flow is simul

  8. Residence Time Distribution Measurement and Analysis of Pilot-Scale Pretreatment Reactors for Biofuels Production: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, D.; Kuhn, E.; Tucker, M.; Stickel, J.; Wolfrum, E.

    2013-06-01

    Measurement and analysis of residence time distribution (RTD) data is the focus of this study where data collection methods were developed specifically for the pretreatment reactor environment. Augmented physical sampling and automated online detection methods were developed and applied. Both the measurement techniques themselves and the produced RTD data are presented and discussed.

  9. Analysis of muon radiography of the Toshiba nuclear critical assembly reactor

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C. L.; Bacon, Jeffery; Borozdin, Konstantin; Fabritius, J. M.; Perry, John; Ramsey, John [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Ban, Yuichiro; Izumi, Mikio; Sano, Yuji; Yoshida, Noriyuki [Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Miyadera, Haruo [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Mizokami, Shinya; Otsuka, Yasuyuki; Yamada, Daichi [Tokyo Electric Power Company, 1-1-3 Uchisaiwai-cho, Chiyoda-ku, Tokyo (Japan); Sugita, Tsukasa; Yoshioka, Kenichi [Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan)

    2014-01-13

    A 1.2 × 1.2 m{sup 2} muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ∼4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.

  10. Comparative analysis of using natural and radiogenic lead as heat-transfer agent in fast reactors

    Science.gov (United States)

    Laas, R. A.; Gizbrekht, R. V.; Komarov, P. A.; Nesterov, V. N.

    2016-06-01

    Fast reactors with lead coolant have several advantages over analogues. Performance can be further improved by replacement of natural composition lead with radiogenic one. Thus, two main issues need to be addressed: induced radioactivity in coolant and efficient neutron multiplication factor in the core will be changed and need to be estimated. To address these issues analysis of the scheme of the nuclear transformations in the lead heat-transfer agent in the process of radiation was carried out. Induced radioactivity of radiogenic and natural lead has been studied. It is shown that replacement of lead affects multiplication factor in a certain way. Application of radiogenic lead can significantly affect reactor operation.

  11. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  12. Nonlinear time-domain soil–structure interaction analysis of embedded reactor structures subjected to earthquake loads

    Energy Technology Data Exchange (ETDEWEB)

    Solberg, Jerome M., E-mail: solberg2@llnl.gov [Methods Development Group, Lawrence Livermore Nat’l Lab, P.O. Box 808, Mailstop L-125, Livermore, CA 94550 (United States); Hossain, Quazi, E-mail: hossain1@llnl.gov [Structural and Applied Mechanics Group, Lawrence Livermore Nat’l Lab, P.O. Box 808, Mailstop L-129, Livermore, CA 94550 (United States); Mseis, George, E-mail: george.mseis@gmail.com [Structural and Applied Mechanics Group, Lawrence Livermore Nat’l Lab, P.O. Box 808, Mailstop L-129, Livermore, CA 94550 (United States)

    2016-08-01

    Highlights: • Derived modified version of Bielak’s SSI method for nonlinear time-domain analysis. • Utilized a Ramberg–Osgood material with parameters that can be fit to EPRI data. • Matched vertically propagating shear wave results from CARES. • Applied this technique to a representative SMR, compared well with SASSI. • The technique is extensible to other material models and nonlinear effects. - Abstract: A generalized time-domain method for soil–structure interaction analysis is developed, based upon an extension of the work of the domain reduction method of Bielak et al. The methodology is combined with the use of a simple hysteretic soil model based upon the Ramberg–Osgood formulation and applied to a notional Small Modular Reactor. These benchmark results compare well (with some caveats) with those obtained by using the industry-standard frequency-domain code SASSI. The methodology provides a path forward for investigation of other sources of nonlinearity, including those associated with the use of more physically-realistic material models incorporating pore-pressure effects, gap opening/closing, the effect of nonlinear structural elements, and 3D seismic inputs.

  13. Cost-utility analysis: Current methodological issues and future perspectives

    Directory of Open Access Journals (Sweden)

    Mark J C Nuijten

    2011-06-01

    Full Text Available The use of cost-effectiveness as final criterion in the reimbursement process for listing of new pharmaceuticals can be questioned from a scientific and policy point of view. There is a lack of consensus on main methodological issues and consequently we may question the appropriateness of the use of cost-effectiveness data in health care decision-making. Another concern is the appropriateness of the selection and use of an incremental cost-effectiveness threshold (Cost/QALY. In this review, we focus mainly on only some key methodological concerns relating to discounting, the utility concept, cost assessment and modelling methodologies. Finally we will consider the relevance of some other important decision criteria, like social values and equity.

  14. Spent fuel acceptance scenarios devoted to shutdown reactors: A preliminary analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wood, T.W.; Plummer, A.M.; Dippold, D.G.; Short, S.M. (Pacific Northwest Lab., Richland, WA (USA); Battelle Memorial Inst., Columbus, OH (USA). Office of Transportation Systems and Planning; Pacific Northwest Lab., Richland, WA (USA))

    1989-10-01

    Spent fuel acceptance schedules and the allocation of federal acceptance capacity among commercial nuclear power reactors have important operational and cost consequences for reactor operators. Alternative allocation schemes were investigated to some extent in DOE's MRS Systems Study. The current study supplements these analyses for a class of acceptance schemes in which the acceptance capacity of the federal radioactive waste management system is allocated principally to shutdown commercial power reactors, and extends the scope of analysis to include considerations of at-reactor cask loading rates. The operational consequences of these schemes for power reactors, as measured in terms of quantity of spent fuel storage requirement above storage pool capacities and number of years of pool operations after last discharge, are estimated, as are the associated utility costs. This study does not attempt to examine the inter-utility equity considerations involved in departures from the current oldest-fuel-first (OFF) allocation rule as specified in the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste.'' In the sense that the alternative allocations are more economically efficient than OFF, however, they approximate the allocations that could result from free exchange of acceptance rights among utilities. Such a process would result in the preservation of inter-utility equity. 13 refs., 9 figs., 9 tabs.

  15. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  16. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  17. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Science.gov (United States)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-09-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  18. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  19. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-12-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  20. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-07-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  1. Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Wosko, Paul; Sundram, S. K.

    2012-10-16

    New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

  2. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  3. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

  4. Attack Methodology Analysis: Emerging Trends in Computer-Based Attack Methodologies and Their Applicability to Control System Networks

    Energy Technology Data Exchange (ETDEWEB)

    Bri Rolston

    2005-06-01

    Threat characterization is a key component in evaluating the threat faced by control systems. Without a thorough understanding of the threat faced by critical infrastructure networks, adequate resources cannot be allocated or directed effectively to the defense of these systems. Traditional methods of threat analysis focus on identifying the capabilities and motivations of a specific attacker, assessing the value the adversary would place on targeted systems, and deploying defenses according to the threat posed by the potential adversary. Too many effective exploits and tools exist and are easily accessible to anyone with access to an Internet connection, minimal technical skills, and a significantly reduced motivational threshold to be able to narrow the field of potential adversaries effectively. Understanding how hackers evaluate new IT security research and incorporate significant new ideas into their own tools provides a means of anticipating how IT systems are most likely to be attacked in the future. This research, Attack Methodology Analysis (AMA), could supply pertinent information on how to detect and stop new types of attacks. Since the exploit methodologies and attack vectors developed in the general Information Technology (IT) arena can be converted for use against control system environments, assessing areas in which cutting edge exploit development and remediation techniques are occurring can provide significance intelligence for control system network exploitation, defense, and a means of assessing threat without identifying specific capabilities of individual opponents. Attack Methodology Analysis begins with the study of what exploit technology and attack methodologies are being developed in the Information Technology (IT) security research community within the black and white hat community. Once a solid understanding of the cutting edge security research is established, emerging trends in attack methodology can be identified and the gap between

  5. Hazard analysis for 300 Area N Reactor Fuel Fabrication and Storage Facilty

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D.J.; Brehm, J.R.

    1994-01-25

    This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. Unlike the Facility hazard classification documented in WHC-SD-NR-HC-004, Hazard Classification for 300 Area N Reactor Fuel Fabrication and Storage Facility, (Huang 1993), which is based on unmitigated consequences, credit is taken in an HA for administrative controls or engineered safety features planned or in place. The HA is the foundation for the accident analysis. The significant event scenarios identified by this HA will be further evaluated in a subsequent accident analysis.

  6. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  7. Adjoint-based uncertainty quantification and sensitivity analysis for reactor depletion calculations

    Science.gov (United States)

    Stripling, Hayes Franklin

    Depletion calculations for nuclear reactors model the dynamic coupling between the material composition and neutron flux and help predict reactor performance and safety characteristics. In order to be trusted as reliable predictive tools and inputs to licensing and operational decisions, the simulations must include an accurate and holistic quantification of errors and uncertainties in its outputs. Uncertainty quantification is a formidable challenge in large, realistic reactor models because of the large number of unknowns and myriad sources of uncertainty and error. We present a framework for performing efficient uncertainty quantification in depletion problems using an adjoint approach, with emphasis on high-fidelity calculations using advanced massively parallel computing architectures. This approach calls for a solution to two systems of equations: (a) the forward, engineering system that models the reactor, and (b) the adjoint system, which is mathematically related to but different from the forward system. We use the solutions of these systems to produce sensitivity and error estimates at a cost that does not grow rapidly with the number of uncertain inputs. We present the framework in a general fashion and apply it to both the source-driven and k-eigenvalue forms of the depletion equations. We describe the implementation and verification of solvers for the forward and ad- joint equations in the PDT code, and we test the algorithms on realistic reactor analysis problems. We demonstrate a new approach for reducing the memory and I/O demands on the host machine, which can be overwhelming for typical adjoint algorithms. Our conclusion is that adjoint depletion calculations using full transport solutions are not only computationally tractable, they are the most attractive option for performing uncertainty quantification on high-fidelity reactor analysis problems.

  8. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  9. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  10. Simulation of a Neutron Noise Analysis Method for the Detection of Reactor Internals Vibration

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of); Song, Seon Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The results were compared against expected hypothesis. This simulation technique developed in C++ programming environment can successfully illustrate the principle of the neutron noise analysis as a vibration monitoring method. The addition of a white noise signal spectrum into the neutron flux data may result in a better coherence analysis. Examination of the phase data on adjacent and opposite flux pairs may be used to determine the vibration mode 3 session. Safety aspect is always highly demanded in any nuclear power plants operation. To achieve a high level of safety, it is desirable to perform preventive measures instead of corrective ones. One of these measures is the monitoring of reactor internals vibration characteristics. Any changes in the vibration signatures indicates an anomaly in the reactor internals. One proven method for this purpose is by analyzing the neutron flux sensed by ex-core detectors around the reactor core. Standards and guides have been written on the proper conduct of this method. The American Society of Mechanical Engineers (ASME) published two similar guides in the ASME OM-S/G-2007 document. Part 5 focuses on specifically monitoring the core support barrel axial preload. Part 23 elaborates on monitoring of reactor internals vibrations. U. S. Nuclear Regulatory Commission (NRC) issued a Regulatory Guide 1.20 on Comprehensive Vibration Assessment Program (CVAP). To understand the principle of neutron noise analysis on vibration monitoring, a simple neutron-transport model was simulated.

  11. An Analysis of the Research Methodology of the Ramirez Study.

    Science.gov (United States)

    Thomas, Wayne P.

    1992-01-01

    Analyzes the political, educational, and technical factors that strongly influenced the Ramirez study of bilingual programs. Enumerates strengths and weaknesses of the study's research methodology, along with implications for decision making in language-minority education. Summarizes defensible conclusions of the study that have not yet been…

  12. An Analysis of the Research Methodology of the Ramirez Study.

    Science.gov (United States)

    Thomas, Wayne P.

    1992-01-01

    Analyzes the political, educational, and technical factors that strongly influenced the Ramirez study of bilingual programs. Enumerates strengths and weaknesses of the study's research methodology, along with implications for decision making in language-minority education. Summarizes defensible conclusions of the study that have not yet been…

  13. Towards a Methodological Improvement of Narrative Inquiry: A Qualitative Analysis

    Science.gov (United States)

    Abdallah, Mahmoud Mohammad Sayed

    2009-01-01

    The article suggests that though narrative inquiry as a research methodology entails free conversations and personal stories, yet it should not be totally free and fictional as it has to conform to some recognized standards used for conducting educational research. Hence, a qualitative study conducted by Russ (1999) was explored as an exemplar…

  14. Biological treatment of a synthetic dairy wastewater in a sequencing batch biofilm reactor: Statistical modeling using optimization using response surface methodology

    Directory of Open Access Journals (Sweden)

    Zinatizadeh A.A.L.

    2011-01-01

    Full Text Available In this study, the interactive effects of initial chemical oxygen demand (CODin, biomass concentration and aeration time on the performance of a lab-scale sequencing batch biofilm reactor (SBBR treating a synthetic dairy wastewater were investigated. The experiments were conducted based on a central composite design (CCD and analyzed using response surface methodology (RSM. The region of exploration for treatment of the synthetic dairy wastewater was taken as the area enclosed by the influent comical oxygen demand (CODin (1000, 3000 and 5000 mg/l, biomass concentration (3000, 5000 and 7000 mg VSS/l and aeration time (2, 8 and 18 h boundaries. Two dependent parameters were measured or calculated as response. These parameters were total COD removal efficiency and sludge volume index (SVI. The maximum COD removal efficiencies (99.5% were obtained at CODin, biomass concentration and aeration time of 5000 mg COD/l, 7000 mg VSS/l and 18 h, respectively. The present study provides valuable information about interrelations of quality and process parameters at different values of the operating variables.

  15. Improving the simultaneous removal of chemical oxygen demand and terephthalic acid in a cross-flow aerobic sludge reactor by using response surface methodology.

    Science.gov (United States)

    Hu, Dong-Xue; Tian, Yu; Chen, Zhao-Bo; Ge, Hui; Cui, Yu-Bo; Ran, Chun-Qiu

    2015-01-01

    Central composite design and response surface methodology (RSM) were implemented to optimize the operational parameters for a cross-flow aerobic sludge reactor (CFASR) in remedying mixed printing and dyeing wastewater (MPDW). The individual and interactive effects of three variables, hydraulic retention time (HRT), pH and sludge loading rate (SLR), on chemical oxygen demand (COD) and terephthalic acid (TA) removal rates were evaluated. For HRT of 15.3-19.8 hours, pH of 7.2-8.1 and SLR of 0.4-0.6 kg chemical oxygen demand (COD) per kg mixed liquor suspended solids per day, COD and TA removal rates of the CFASR exceeded 85% and 90%, respectively. The check experiment revealed that the effluent from the optimized CFASR was stable below the limitation of 100 mg COD/L and the TA concentration decreased by 6.0% compared to the usual CFASR. The results verified that the RSM was useful for optimizing the operation parameters of the CFASR in remedying MPDW.

  16. Optimal Conditions for Continuous Immobilization of Pseudozyma hubeiensis (Strain HB85A Lipase by Adsorption in a Packed-Bed Reactor by Response Surface Methodology

    Directory of Open Access Journals (Sweden)

    Roberta Bussamara

    2012-01-01

    Full Text Available This study aimed to develop an optimal continuous process for lipase immobilization in a bed reactor in order to investigate the possibility of large-scale production. An extracellular lipase of Pseudozyma hubeiensis (strain HB85A was immobilized by adsorption onto a polystyrene-divinylbenzene support. Furthermore, response surface methodology (RSM was employed to optimize enzyme immobilization and evaluate the optimum temperature and pH for free and immobilized enzyme. The optimal immobilization conditions observed were 150 min incubation time, pH 4.76, and an enzyme/support ratio of 1282 U/g support. Optimal activity temperature for free and immobilized enzyme was found to be 68°C and 52°C, respectively. Optimal activity pH for free and immobilized lipase was pH 4.6 and 6.0, respectively. Lipase immobilization resulted in improved enzyme stability in the presence of nonionic detergents, at high temperatures, at acidic and neutral pH, and at high concentrations of organic solvents such as 2-propanol, methanol, and acetone.

  17. Acidolysis of terebinth fruit oil with palmitic and caprylic acids in a recirculating packed bed reactor: optimization using response surface methodology

    Energy Technology Data Exchange (ETDEWEB)

    Koçak, D.; Keskin, H.; Fadiloglu, S.; Gögüs, F.

    2016-07-01

    The acidolysis reaction of terebinth fruit oil with caprylic and palmitic acid has been investigated. The reaction was catalyzed by lipase (Lipozyme IM from Rhizomucormiehei) and carried out in recirculating packed bed reactor. The effects of reaction parameters have been analyzed using response surface methodology. Reaction time (3.5–6.5 h), enzyme load (10–20%), substrate flow rate (4–8 mL·min−1 ) and substrate mole ratios (Terebinth oil : Palmitic acid : Caprylic acid, 1:1.83:1.22–1:3.07:2.05) were evaluated. The optimum reaction conditions were 5.9 h reaction time, 10% enzyme load, 4 mL·min−1 substrate flow rate and 1:3.10:2.07 substrate mole ratio. The structured lipid obtained at these optimum conditions had 52.23% desired triacylglycerols and a lower caloric value than that of terebinth fruit oil. The melting characteristics and microstructure of the structured lipid were similar to those of commercial margarine fat extracts. The results showed that the structured lipid had the highest oxidative stability among the studied fats. (Author)

  18. 3RD WP PROBABILISTIC CRITICALITY ANALYSIS: METHODOLOGY FOR BASKET DEGRADATION WITH APPLICANTION TO COMMERICAL SNF

    Energy Technology Data Exchange (ETDEWEB)

    P. Goulib

    1997-09-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to describe the latest version of the probabilistic criticality analysis methodology and its application to the entire commercial waste stream of commercial pressurized water reactor (PWR) spent nuclear fuel (SNF) expected to be emplaced in the repository. The purpose of this particular application is to evaluate the 21 assembly PWR absorber plate waste package (WP) with respect to degraded mode criticality performance. The degradation of principal concern is the borated stainless steel absorber plates which are part of the waste package basket and which constitute a major part of the waste package criticality control. The degradation (corrosion, dissolution) of this material will result in the release of most of the boron from the waste package and increase the possibility of criticality. The results of this evaluation will be expressed in terms of the fraction of the PWR SNF which can exceed a given k{sub eff}, as a function of time and the peak value of that fraction over a time period up to several hundred thousand years. The ultimate purpose of this analysis is to support the waste package design which defines waste packages to cover a range of SNF characteristics. In particular, with respect to PWR criticality the current categories are: (1) no specific criticality control material, (2) borated stainless steel plates in the waste package basket, and (3) zirconium clad boron carbide control rods (Ref. 5.4). The results of this analysis will indicate the coverage provided by the first two categories. With these results, this study will provide the first quantitative estimate of the benefit expected from the control measure consisting of borated stainless steel plates. This document is the third waste package probabilistic criticality analysis. The first two (Ref. 5.12 for the first and Ref. 5.15 for the second) analyses were based primarily on the

  19. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States)

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  20. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  1. Analysis of regulatory structure for a potential fusion reactor industry

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The report is divided into eight sections. The preface describes the authors of the report, the methodology used in its preparation, and some basic legal terms. The summary describes the principal features of the proposed regulatory system and also includes two flow charts comparing our model with present NRC practices and a summary table briefly outlining the reasoning behind our recommendations. The main body of the report is divided into six sections. This part of the report discusses the existing federal and state programs for regulating electric energy, describes NRC operations and the criticisms of that agency, discusses the features of our proposed regulatory model, recommends certain steps for implementing the proposed model, and states the conclusions of the report.

  2. PHISICS TOOLKIT: MULTI-REACTOR TRANSMUTATION ANALYSIS UTILITY - MRTAU

    Energy Technology Data Exchange (ETDEWEB)

    Andrea Alfonsi; Cristian Rabiti; Aaron S. Epiney; Yaqi Wang; Joshua Cogliati

    2012-04-01

    The principal idea of this paper is to present the new capabilities available in the PHISICS toolkit, connected with the implementation of the depletion code MRTAU, a generic depletion/ decay/burn-up code developed at the Idaho National Laboratory. It is programmed in a modular structure and modern FORTRAN 95/2003. The code tracks the time evolution of the isotopic concentration of a given material accounting for nuclear reaction happening in presence of neutron flux and also due to natural decay. MRTAU has two different methods to perform the depletion calculation, in order to let the user choose the best one respect his needs. Both the methodologies and some significant results are reported in this paper.

  3. Software for neutron activation analysis at reactor IBR-2, FLNP, JINR

    CERN Document Server

    Zlokazov, V B

    2004-01-01

    A Delphi program suite, developed for processing gamma-spectra of induced activity of nuclei, obtained from the neutron activation measurements at the reactor IBR-2, FLNF, JINR, is reported. This suite contains components, intended for carrying out all the operations of the analysis cycle, starling with a data acquisition program for gamma -spectrometers Gamma (written in C++ Builder) and including Delphi programs for steps of the analysis. (6 refs).

  4. On the extension of modern best-estimate plus uncertainy methodologies to future fast reactor and advanced fuel licensing - initial evaluation of issues

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Mcclure, Patrick R [Los Alamos National Laboratory

    2009-01-01

    and has precedence in the NRC. The method is generically referred to as a 'Best Estimate plus Uncertainty' approach (BE+U), since the goal of the methodology is to compare the model value (best estimate) plus any uncertainty to a figure of merit like cladding temperature. The challenges for extending the BE+U (1) method for fuel qualification for an Advanced Reactor Fuel are driven by: schedule, the need for data, the data sufficiency, the identification of important phenomenon, the process of validation (with focus on the multi-scale model), and the need to produce and extended best estimate plus uncertainty methodology. This paper examines these issues an offers up a proposed set of methods that extend the current BE+U methodology address most if not all of these challenges.

  5. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  6. NUMERICAL ANALYSIS OF THERMAL STRATIFICATION IN THE UPPER PLENUM OF THE MONJU FAST REACTOR

    Directory of Open Access Journals (Sweden)

    SEOK-KI CHOI

    2013-04-01

    Full Text Available A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (∼300 seconds. However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

  7. Full-Envelope Launch Abort System Performance Analysis Methodology

    Science.gov (United States)

    Aubuchon, Vanessa V.

    2014-01-01

    The implementation of a new dispersion methodology is described, which dis-perses abort initiation altitude or time along with all other Launch Abort System (LAS) parameters during Monte Carlo simulations. In contrast, the standard methodology assumes that an abort initiation condition is held constant (e.g., aborts initiated at altitude for Mach 1, altitude for maximum dynamic pressure, etc.) while dispersing other LAS parameters. The standard method results in large gaps in performance information due to the discrete nature of initiation conditions, while the full-envelope dispersion method provides a significantly more comprehensive assessment of LAS abort performance for the full launch vehicle ascent flight envelope and identifies performance "pinch-points" that may occur at flight conditions outside of those contained in the discrete set. The new method has significantly increased the fidelity of LAS abort simulations and confidence in the results.

  8. Anaerobic digestion of manure and mixture of manure with lipids: biogas reactor performance and microbial community analysis

    DEFF Research Database (Denmark)

    Mladenovska, Zuzana; Dabrowski, Slawomir; Ahring, Birgitte Kiær

    2003-01-01

    and a higher removal of VS than the reactor treating manure. Microbial population analysis done by cultivation - most probable number (MPN) test and specific methanogenic activity (SMA) measurement, revealed higher MPN and increased SMA of methanogenic populations of biomass from the reactor codigesting manure...

  9. Safety analysis of 5 MW IEAR-1 reactor; Analise de seguranca do reator IEA-R1 a 5 MW

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Antonio T. e; Maprelian, Eduardo; Rodrigues, Antonio C.I.; Cabral, Eduardo L.L.; Molnary, Leslie de; Mesquita, Ricardo N.; Mendonca, Arlindo G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores. E-mail: teixeira@net.ipen.br

    2000-07-01

    This paper presents the methods and procedures utilized in the safety analysis of IEA-R1 research reactor. Four postulated accidents are quantitatively analyzed, being the fuel channel blockage accident considered as the Maximum Credible Accident for the reactor. The potential accident consequences and the criteria for radiological doses acceptance are evaluated and discussed. (author)

  10. Population Analysis: A Methodology for Understanding Populations in COIN Environments

    Science.gov (United States)

    2008-12-01

    contributions to the field. Emile Durkheim (1858-1917) is described as one of the founders of sociology. In the earliest days of sociology, thinkers...such as Durkheim , attempted to apply a scientific methodological approach to understanding societal change. During this era, philosophers such as... Durkheim , Max Weber and Karl Marx each contributed to the field in their attempts to explain the world around them. At a very basic level, this

  11. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

  12. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  13. French Epistemology and its Revisions: Towards a Reconstruction of the Methodological Position of Foucaultian Discourse Analysis

    OpenAIRE

    Rainer Diaz-Bone

    2007-01-01

    This article reconstructs epistemology in the tradition of Gaston BACHELARD as one of the main foundations of the methodology of FOUCAULTian discourse analysis. Foundational concepts and the methodological approach of French epistemology are one of the continuities in the work of Michel FOUCAULT. BACHELARDian epistemology (and of his successor Georges CANGUILHEM) can be used for the reconstruction of the FOUCAULTian methodology and it can also be used to instruct the practices of FOUCAULTian ...

  14. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  15. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  16. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  17. Analysis of Postulated Core Meltdown of an SRP Reactor - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Durant, W.S.; Brown, R.J.

    1970-10-01

    An analysis was made to determine the consequences of a postulated accident in which the core of a Savannah River Plant reactor melts down following the loss of coolant. The study was made to determine (1) the potential damage to the reactor building that could impair its integrity for confining activity and (2) the need for additional facilities to prevent the activity confinement system from being overheated by the decay heat in the debris. A preliminary report on this analysis was issued previously. The sequence of events during and following the loss of coolant has now been studied in more detail, and a computer program has been written and used to investigate transient heating effects. This is the final report of the analysis and presents the conclusions.

  18. Progress report on neutron activation analysis at Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tuan, Nguyen Ngoc [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)

  19. INTEGRATED METHODOLOGY FOR PRODUCT PLANNING USING MULTI CRITERIA ANALYSIS

    Directory of Open Access Journals (Sweden)

    Tarun Soota

    2016-09-01

    Full Text Available Integrated approach to multi-criteria decision problems is proposed using quality function deployment and analytical network process. The objective of the work is to rationalize and improve the method of analyzing and interpreting customer needs and technical requirements. The methodology is used to determine, prioritize engineering requirements based on customer needs for development of best product. Framework allows decision maker to decompose a complex problem in a hierarchical structure to show relationship between objective and criteria. Multi-criteria decision modeling is used for extending the hierarchy process to both dependence and feedback. A case study on bikes is presented for the proposed model.

  20. Development of seismic sloshing analysis method of liquid coolant sodium in the KALIMER reactor vessel including several cylindrical components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong

    2000-11-01

    It is important to establish a highly accurate technique of evaluating the sloshing behavior of liquid sodium coolant during earthquake for structural integrity of KALIMER reactor vessel and internals. The analysis procedure of sloshing behaviors is established using finite element computer program ANSYS, and the effectiveness of the procedure is confirmed by comparison with theoretical and experimental results in the literature. The analysis results agree well with experimental ones. Based on the procedure, the sloshing characteristics of liquid sodium coolant in the KALIMER reactor vessel including reactor internal components are evaluated. The maximum response height of sodium free surface at the reactor vessel is about 55cm when subjected to horizontal safe shutdown earthquake (SSE) of 0.3g for seismically isolated reactor building.

  1. A study on the core analysis methodology for SMART CEA ejection accident-I

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Lee, Chung Chan; Kim, Kyo Yoon; Cho, Byung Oh

    1999-04-01

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs.

  2. USTED: a methodology for a quantitative analysis of land use scenarios.

    NARCIS (Netherlands)

    Stoorvogel, J.J.; Schipper, R.A.; Jansen, D.M.

    1995-01-01

    Scenarios are a useful tool to study the effects of changes in the socio-economic and/or biophysical environment on agricultural land use. For the analysis of these scenarios an operational methodology, USTED, has been developed. The methodology is based on a linear programming model in combination

  3. CFD analysis and flow model reduction for surfactant production in helix reactor

    Directory of Open Access Journals (Sweden)

    Nikačević N.M.

    2015-01-01

    Full Text Available Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD simulations. Non-reactive flow is simulated, though viscosity is treated as variable in the direction of flow, as it increases during the reaction. The design and operating parameters (reactor diameter, number of coils and inlet velocity are varied in CFD simulations, in order to examine the effects on the flow pattern. Given that 3D simulations are not practical for fast computations needed for optimization, scale-up and control, CFD flow model is reduced to one-dimensional axial dispersion (AD model with spatially variable dispersion coefficient. Dimensionless dispersion coefficient (Pe is estimated under different conditions and results are analyzed. Finally, correlation which relates Pe number with Reynolds number and number of coils from the reactor entrance is proposed for the particular reactor application and conditions.

  4. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N.J.H.

    1966-12-15

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

  5. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Junli Gou

    2009-01-01

    Full Text Available A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS, which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS, the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

  6. Analysis of the key enzymes of butyric and acetic acid fermentation in biogas reactors.

    Science.gov (United States)

    Gabris, Christina; Bengelsdorf, Frank R; Dürre, Peter

    2015-09-01

    This study aimed at the investigation of the mechanisms of acidogenesis, which is a key process during anaerobic digestion. To expose possible bottlenecks, specific activities of the key enzymes of acidification, such as acetate kinase (Ack, 0.23-0.99 U mg(-1) protein), butyrate kinase (Buk, < 0.03 U mg(-1) protein) and butyryl-CoA:acetate-CoA transferase (But, 3.24-7.64 U mg(-1) protein), were determined in cell free extracts of biogas reactor content from three different biogas reactors. Furthermore, the detection of Ack was successful via Western blot analysis. Quantification of corresponding functional genes encoding Buk (buk) and But (but) was not feasible, although an amplification was possible. Thus, phylogenetic trees were constructed based on respective gene fragments. Four new clades of possible butyrate-producing bacteria were postulated, as well as bacteria of the genera Roseburia or Clostridium identified. The low Buk activity was in contrast to the high specific But activity in the analysed samples. Butyrate formation via Buk activity does barely occur in the investigated biogas reactor. Specific enzyme activities (Ack, Buk and But) in samples drawn from three different biogas reactors correlated with ammonia and ammonium concentrations (NH₃ and NH₄(+)-N), and a negative dependency can be postulated. Thus, high concentrations of NH₃ and NH₄(+)-N may lead to a bottleneck in acidogenesis due to decreased specific acidogenic enzyme activities.

  7. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, T. M. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  8. A reactor for high-temperature pyrolysis and oxygen isotopic analysis of cellulose via induction heating.

    Science.gov (United States)

    Evans, Michael N

    2008-07-01

    A reactor for converting cellulose into carbon monoxide for subsequent oxygen isotopic analysis via continuous flow isotope ratio mass spectrometry is described. The system employs an induction heater to produce temperatures >or=1500 degrees C within a molybdenum foil crucible positioned by boron nitride (BN) spacers within a quartz outer sleeve. For samples of a homogeneous working standard cellulose between 300 and 400 microg in size, the blank/signal ratio is <5%, and the long-term precision is 0.30 per thousand (N = 232). For samples of 30 to 100 microg in size, a gas pressure sintered silicon nitride (Si(3)N(4)) outer sleeve replaces the quartz sleeve, the BN spacers are not used, and 6.0-grade carrier He must be used to minimize the blank signal. With these modifications a blank/sample ratio of <5% and long-term precision of 0.30 per thousand (N = 144) are obtained. These results are similar to those achieved using standard high-temperature furnaces, but the reactor is simpler to pack, the system is more economical to run, and samples as small as 30 microg cellulose may be measured. For both reactors memory is significant in the subsequent sample and is believed to be due to exchange with reactor oxygen at temperatures above 1000 degrees C. Further applications might include online preparation of other materials requiring temperatures of 1500-2600 degrees C.

  9. Analysis of the key enzymes of butyric and acetic acid fermentation in biogas reactors

    Science.gov (United States)

    Gabris, Christina; Bengelsdorf, Frank R; Dürre, Peter

    2015-01-01

    This study aimed at the investigation of the mechanisms of acidogenesis, which is a key process during anaerobic digestion. To expose possible bottlenecks, specific activities of the key enzymes of acidification, such as acetate kinase (Ack, 0.23–0.99 U mg−1 protein), butyrate kinase (Buk, biogas reactor content from three different biogas reactors. Furthermore, the detection of Ack was successful via Western blot analysis. Quantification of corresponding functional genes encoding Buk (buk) and But (but) was not feasible, although an amplification was possible. Thus, phylogenetic trees were constructed based on respective gene fragments. Four new clades of possible butyrate-producing bacteria were postulated, as well as bacteria of the genera Roseburia or Clostridium identified. The low Buk activity was in contrast to the high specific But activity in the analysed samples. Butyrate formation via Buk activity does barely occur in the investigated biogas reactor. Specific enzyme activities (Ack, Buk and But) in samples drawn from three different biogas reactors correlated with ammonia and ammonium concentrations (NH3 and NH4+-N), and a negative dependency can be postulated. Thus, high concentrations of NH3 and NH4+-N may lead to a bottleneck in acidogenesis due to decreased specific acidogenic enzyme activities. PMID:26086956

  10. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600.

  11. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2001-07-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  12. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  13. A methodological comparison of customer service analysis techniques

    Science.gov (United States)

    James Absher; Alan Graefe; Robert Burns

    2003-01-01

    Techniques used to analyze customer service data need to be studied. Two primary analysis protocols, importance-performance analysis (IP) and gap score analysis (GA), are compared in a side-by-side comparison using data from two major customer service research projects. A central concern is what, if any, conclusion might be different due solely to the analysis...

  14. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  15. Grinding analysis of Indian coal using response surface methodology

    Institute of Scientific and Technical Information of China (English)

    Twinkle Singh; Aishwarya Awasthi; Pranjal Tripathi; Shina Gautam; Alok Gautam

    2016-01-01

    The present work discusses a systematic approach to model grinding parameters of coal in a ball mill.A three level Box-Behnken design combined with response surface methodology using second order model was applied to the experiments done according to the model requirement.Three parameters ball charge (numbers 10-20),coal content (100-200 g) and the grinding time (4-8 min) were chosen for the experiments as well as for the modeling work.Coal fineness is defined as the d80 (80 % passing size).A quadratic model was developed to show the effect of parameters and their interaction with fineness of the product.Three different sizes (4,1 and 0.65 mm) of Indian coal were used.The model equations for each fraction were developed and different sets of experiments were performed.The predicted values of the fineness of coal were in good agreement with the experimental results (R2 values of d80 varies between 0.97 and 0.99).Fine size of three different coal sizes were obtained with larger ball charge with less grinding time and less solid content.This work represents the efficient use of response surface methodology and the Box-Behnken design use for grinding of Indian coal.

  16. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  17. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  18. Integrated physics analysis of plasma start-up scenario of helical reactor FFHR-d1

    Science.gov (United States)

    Goto, T.; Miyazawa, J.; Sakamoto, R.; Seki, R.; Suzuki, C.; Yokoyama, M.; Satake, S.; Sagara, A.; The FFHR Design Group

    2015-06-01

    1D physics analysis of the plasma start-up scenario of the large helical device (LHD)-type helical reactor FFHR-d1 was conducted. The time evolution of the plasma profile is calculated using a simple model based on the LHD experimental observations. A detailed assessment of the magnetohydrodynamic equilibrium and neo-classical energy loss was conducted using the integrated transport analysis code TASK3D. The robust controllability of the fusion power was confirmed by feedback control of the pellet fuelling and a simple staged variation of the external heating power with a small number of simple diagnostics (line-averaged electron density, edge electron density and fusion power). A baseline operation control scenario (plasma start-up and steady-state sustainment) of the FFHR-d1 reactor for both self-ignition and sub-ignition operation modes was demonstrated.

  19. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.

  20. Analysis, evaluation, and optimization of kinetic parameters for performance appraisal and design of UASB reactors.

    Science.gov (United States)

    Bhunia, Puspendu; Ghangrekar, M M

    2008-05-01

    Studies have been undertaken to explore the applicability of different kinetic models for the performance appraisal of upflow anaerobic sludge blanket (UASB) reactors treating wastewater in the range of 300-4000 mg COD/l. Three kinetic models namely, Monod, Grau second-order, and Haldane model are considered for the analysis. Both linear and nonlinear regressions have been performed to examine the best-fit among the kinetic models. In this process, five error analysis methods have been used to analyze the data. Apart from optimization of kinetic coefficients with minimization of associated errors, prediction of effluent COD has also been undertaken to verify the applicability of kinetic models. In both the cases, Grau second-order model is found to be the best class of fit for wide range of data sets in UASB reactor.

  1. Application of response surface methodology (RSM) for optimisation of COD, NH3-N and 2,4-DCP removal from recycled paper wastewater in a pilot-scale granular activated carbon sequencing batch biofilm reactor (GAC-SBBR).

    Science.gov (United States)

    Muhamad, Mohd Hafizuddin; Sheikh Abdullah, Siti Rozaimah; Mohamad, Abu Bakar; Abdul Rahman, Rakmi; Hasan Kadhum, Abdul Amir

    2013-05-30

    In this study, the potential of a pilot-scale granular activated carbon sequencing batch biofilm reactor (GAC-SBBR) for removing chemical oxygen demand (COD), ammoniacal nitrogen (NH3-N) and 2,4-dichlorophenol (2,4-DCP) from recycled paper wastewater was assessed. For this purpose, the response surface methodology (RSM) was employed, using a central composite face-centred design (CCFD), to optimise three of the most important operating variables, i.e., hydraulic retention time (HRT), aeration rate (AR) and influent feed concentration (IFC), in the pilot-scale GAC-SBBR process for recycled paper wastewater treatment. Quadratic models were developed for the response variables, i.e., COD, NH3-N and 2,4-DCP removal, based on the high value (>0.9) of the coefficient of determination (R(2)) obtained from the analysis of variance (ANOVA). The optimal conditions were established at 750 mg COD/L IFC, 3.2 m(3)/min AR and 1 day HRT, corresponding to predicted COD, NH3-N and 2,4-DCP removal percentages of 94.8, 100 and 80.9%, respectively.

  2. Remarks on boiling water reactor stability analysis. Pt. 1. Theory and application of bifurcation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hurtado, Antonio [Technische Univ. Dresden (Germany). Chair of Hydrogen and Nuclear Energy; Schuster, Roland [Kernkraftwerk Brunsbuettel GmbH und Co. oHG, Brunsbuettel (Germany); Lukas, Bernard [EnBW Kernkraft GmbH, Philippsburg (Germany). Kernkraftwerk Philippsburg; Aguirre, Carlos [Kernkraftwerk Leibstadt AG, Aargau (Switzerland); Hennig, Dieter

    2012-11-15

    Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)

  3. Fluid Dynamic Analysis of Shaken and Mechanically Stirred Reactors by Means of Optical Techniques

    OpenAIRE

    Pieralisi, Irene

    2016-01-01

    This research aims at deepening the knowledge of the hydrodynamics developing within some specific mixing system configurations. In particular, a standard geometry stirred vessel, an unbaffled stirred vessel of unconventional geometry and a shaken bioreactor are studied. Optical techniques (LDA, PIV) are employed in order to obtain detailed information on the mean and turbulent characteristics of the flow developing within the different lab-scale reactors, while subsequent data analysis a...

  4. A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

    Science.gov (United States)

    Menke, M. M.; Judd, B. R.

    1973-01-01

    The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

  5. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  6. Analysis of Requirement Engineering Processes, Tools/Techniques and Methodologies

    Directory of Open Access Journals (Sweden)

    Tousif ur Rehman

    2013-02-01

    Full Text Available Requirement engineering is an integral part of the software development lifecycle since the basis for developing successful software depends on comprehending its requirements in the first place. Requirement engineering involves a number of processes for gathering requirements in accordance with the needs and demands of users and stakeholders of the software product. In this paper, we have reviewed the prominent processes, tools and technologies used in the requirement gathering phase. The study is useful to perceive the current state of the affairs pertaining to the requirement engineering research and to understand the strengths and limitations of the existing requirement engineering techniques. The study also summarizes the best practices and how to use a blend of the requirement engineering techniques as an effective methodology to successfully conduct the requirement engineering task. The study also highlights the importance of security requirements as though they are part of the non-functional requirement, yet are naturally considered fundamental to secure software development.

  7. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

    Directory of Open Access Journals (Sweden)

    Ordóñez Ródenas José

    2016-01-01

    Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.

  8. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Larry Lorin

    2000-05-01

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  9. Business analysis methodology in telecommunication industry - the research based on the grounded theory

    National Research Council Canada - National Science Library

    Hana Nenickova

    2013-01-01

      The objective of this article is to present the grounded theory using in the qualitative research as a basis to build a business analysis methodology for the implementation of information systems...

  10. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 5. Development and application of reactor analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Kenji; Hazama, Taira; Chiba, Go; Ohki, Shigeo; Ishikawa, Makoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    In the core design of fast reactors (FRs), it is very important to improve the prediction accuracy of the nuclear characteristics for both reducing cost and ensuring reliability of FR plants. A nuclear reactor analysis code system for FRs has been developed by the Japan Nuclear Cycle Development Institute (JNC). This paper describes the outline of the calculation models and methods in the system consisting of several analysis codes, such as the cell calculation code CASUP, the core calculation code TRITAC and the sensitivity analysis code SAGEP. Some examples of verification results and improvement of the design accuracy are also introduced based on the measurement data from critical assemblies, e.g, the JUPITER experiment (USA/Japan), FCA (Japan), MASURCA (France), and BFS (Russia). Furthermore, application fields and future plans, such as the development of new generation nuclear constants and applications to MA{center_dot}FP transmutation, are described. (author)

  11. 75 FR 13610 - Office of New Reactors; Interim Staff Guidance on Implementation of a Seismic Margin Analysis for...

    Science.gov (United States)

    2010-03-22

    ... New Reactors Based on Probabilistic Risk Assessment AGENCY: Nuclear Regulatory Commission (NRC... Probabilistic Risk Assessment,'' (Agencywide Documents Access and Management System (ADAMS) Accession No..., ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' issued...

  12. Molecular analysis of the biomass of a fluidized bed reactor treating synthetic vinasse at anaerobic and micro-aerobic conditions

    NARCIS (Netherlands)

    Rodriguez, E.; Lopes, A.; Fdz-Polanco, M.; Stams, A.J.M.; Garcia Encina, P.A.

    2012-01-01

    The microbial communities (Bacteria and Archaea) established in an anaerobic fluidized bed reactor used to treat synthetic vinasse (betaine, glucose, acetate, propionate, and butyrate) were characterized by denaturing gradient gel electrophoresis (DGGE) and phylogenetic analysis. This study was focu

  13. Uncertainty Analysis of Light Water Reactor Fuel Lattices

    Directory of Open Access Journals (Sweden)

    C. Arenas

    2013-01-01

    Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

  14. REACTOR ANALYSIS AND VIRTUAL CONTROL ENVIRONMENT (RAVEN) FY12 REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Cristian Rabiti; Andrea Alfonsi; Joshua Cogliati; Diego Mandelli; Robert Kinoshita

    2012-09-01

    RAVEN is a complex software tool that will have tasks spanning from being the RELAP-7 user interface, to using RELAP-7 to perform Risk Informed Safety Characterization (RISMC), and to controlling RELAP-7 calculation execution. The goal of this document is to: 1. Highlight the functional requirements of the different tasks of RAVEN 2. Identify shared functions that could be aggregate in modules so to obtain a minimal software redundancy and maximize software utilization. RAVEN is in fact a software framework that will allow exploiting the following functionalities: • Derive and actuate the control logic required to: o Simulate the plant control system o Simulate the operator (procedure guided) actions o Perform Monte Carlo sampling of random distributed events o Perform event three based analysis • Provide a GUI to: o Input a plant description to RELAP-7 (component, control variable, control parameters) o Concurrent monitoring of Control Parameters o Concurrent alteration of control parameters • Provide Post Processing data mining capability based on o Dimensionality reduction o Cardinality reduction In this document it will be shown how an appropriate mathematical formulation of the control logic and probabilistic analysis leads to have most of the software infrastructure leveraged between the two main tasks. Further, this document will go through the development accomplished this year, including simulation results, and priorities for the next years development

  15. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  16. The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, C.-W., E-mail: d937121@oz.nthu.edu.t [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Wang, J.-R.; Lin, H.-T. [Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Cheng, S.-C. [Department of Nuclear Engineering, Taiwan Power Company, No. 242, Sec. 3, Roosevelt Rd., Taipei City 10016, Taiwan (China)

    2010-10-15

    A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using

  17. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  18. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating 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. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  19. Development Risk Methodology for Whole Systems Trade Analysis

    Science.gov (United States)

    2016-08-01

    potential risks, and estimated development times associated with each technology. The established WSTAT framework utilizes elicitation techniques with...and stakeholder value in order to inform and potentially influence requirements documents and associated specifications. The WSTA tool (WSTAT) can...exploitation, requirements definition, early cost informed trades, requirements analysis, Analysis of Alternatives (AoA), contractor trades, and technology

  20. Adapting Job Analysis Methodology to Improve Evaluation Practice

    Science.gov (United States)

    Jenkins, Susan M.; Curtin, Patrick

    2006-01-01

    This article describes how job analysis, a method commonly used in personnel research and organizational psychology, provides a systematic method for documenting program staffing and service delivery that can improve evaluators' knowledge about program operations. Job analysis data can be used to increase evaluators' insight into how staffs…

  1. A Novel Methodology for Thermal Analysis & 3-Dimensional Memory Integration

    CERN Document Server

    Cherian, Annmol; Jose, Jemy; Pangracious, Vinod; 10.5121/ijait.2011.1403

    2011-01-01

    The semiconductor industry is reaching a fascinating confluence in several evolutionary trends that will likely lead to a number of revolutionary changes in the design, implementation, scaling, and the use of computer systems. However, recently Moore's law has come to a stand-still since device scaling beyond 65 nm is not practical. 2D integration has problems like memory latency, power dissipation, and large foot-print. 3D technology comes as a solution to the problems posed by 2D integration. The utilization of 3D is limited by the problem of temperature crisis. It is important to develop an accurate power profile extraction methodology to design 3D structure. In this paper, design of 3D integration of memory is considered and hence the static power dissipation of the memory cell is analysed in transistor level and is used to accurately model the inter-layer thermal effects for 3D memory stack. Subsequently, packaging of the chip is considered and modelled using an architecture level simulator. This modelli...

  2. The XMM Cluster Survey: X-ray analysis methodology

    CERN Document Server

    Lloyd-Davies, E J; Hosmer, Mark; Mehrtens, Nicola; Davidson, Michael; Sabirli, Kivanc; Mann, Robert G; Hilton, Matt; Liddle, Andrew R; Viana, Pedro T P; Campbell, Heather C; Collins, Chris A; Dubois, E Naomi; Freeman, Peter; Hoyle, Ben; Kay, Scott T; Kuwertz, Emma; Miller, Christopher J; Nichol, Robert C; Sahlen, Martin; Stanford, S Adam; Stott, John P

    2010-01-01

    The XMM Cluster Survey (XCS) is a serendipitous search for galaxy clusters using all publicly available data in the XMM- Newton Science Archive. Its main aims are to measure cosmological parameters and trace the evolution of X-ray scaling relations. In this paper we describe the data processing methodology applied to the 5776 XMM observations used to construct the current XCS source catalogue. A total of 3669 > 4-{\\sigma} cluster candidates with >50 background-subtracted X-ray counts are extracted from a total non-overlapping area suitable for cluster searching of 410 deg^2 . Of these, 1022 candidates are detected with >300 X-ray counts, and we demonstrate that robust temperature measurements can be obtained down to this count limit. We describe in detail the automated pipelines used to perform the spectral and surface brightness fitting for these sources, as well as to estimate redshifts from the X-ray data alone. A total of 517 (126) X-ray temperatures to a typical accuracy of <40 (<10) per cent have ...

  3. Recent Methodologies for Creep Deformation Analysis and Its Life Prediction

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woo-Gon; Park, Jae-Young; Iung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To design the high-temperature creeping materials, various creep data are needed for codification, as follows: i) stress vs. creep rupture time for base metals and weldments (average and minimum), ii) stress vs. time to 1% total strain (average), iii) stress vs. time to onset of tertiary creep (minimum), and iv) constitutive eqns. for conducting time- and temperature- dependent stress-strain (average), and v) isochronous stress-strain curves (average). Also, elevated temperature components such as those used in modern power generation plant are designed using allowable stress under creep conditions. The allowable stress is usually estimated on the basis of up to 10{sup 5} h creep rupture strength at the operating temperature. The master curve of the “sinh” function was found to have a wider acceptance with good flexibility in the low stress ranges beyond the experimental data. The proposed multi-C method in the LM parameter revealed better life prediction than a single-C method. These improved methodologies can be utilized to accurately predict the long-term creep life or strength of Gen-IV nuclear materials which are designed for life span of 60 years.

  4. Montecarlo simulation for a new high resolution elemental analysis methodology

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa S, Rodolfo; Brusa, Daniel; Riveros, Alberto [Universidad de La Frontera, Temuco (Chile). Facultad de Ingenieria y Administracion

    1996-12-31

    Full text. Spectra generated by binary, ternary and multielement matrixes when irradiated by a variable energy photon beam are simulated by means of a Monte Carlo code. Significative jumps in the counting rate are shown when the photon energy is just over the edge associated to each element, because of the emission of characteristic X rays. For a given associated energy, the net height of these jumps depends mainly on the concentration and of the sample absorption coefficient. The spectra were obtained by a monochromatic energy scan considering all the emitted radiation by the sample in a 2{pi} solid angle, associating a single multichannel spectrometer channel to each incident energy (Multichannel Scaling (MCS) mode). The simulated spectra were made with Monte Carlo simulation software adaptation of the package called PENELOPE (Penetration and Energy Loss of Positrons and Electrons in matter). The results show that it is possible to implement a new high resolution spectroscopy methodology, where a synchrotron would be an ideal source, due to the high intensity and ability to control the energy of the incident beam. The high energy resolution would be determined by the monochromating system and not by the detection system and not by the detection system, which would basicalbe a photon counter. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Doh; Kil, Chung Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    Data testing of ENDF/B-VI.2 was performed and ACE-format continuous point-wise cross section library from ENDF/B-VI.2 for MCNP was validated through CSEWG benchmark and power plant mockup experiments. The calculated k-effective of ORNL-1, -2, -3, -4 and -10 with ENDF/B-VI are low by about 0.5% but those of L-7, -8, -9, -10 and -11 show good agreement with experiments. Overall results for uranium core with ENDF/B-VI is low in critically than with ENDF/B-V. The calculated results with ENDF/B-VI for PNL-6 {approx} 12 of plutonium core and PNL-30 {approx} 35 of mixed oxide core show good agreement with the experiments. The results of critically calculation for fast core benchmark do not show large difference between ENDF/B-VI and -V. But the calculated results of reaction rate ratio with ENDF/B-VI are improved, compared with ENDF/B-V. The calculated power distribution for VENUS PWR mockup core and typical BWR core of GE with both of ENDF/B-VI and -V agree well with measured values. From the above results, newly generated MCNP library from ENDF/B-VI is useful for nuclear and shielding design and analysis. 5 figs, 13 tabs, 11 refs. (Author).

  6. Seismic Analysis Issues in Design Certification Applications for New Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, M.; Morante, R.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  7. Low-cost activation analysis at small research reactors

    CERN Document Server

    Westphal, G P; Lemmel, H; Niedermaier, M R; Joestl, K; Schröder, P; Böck, H H; Schachner, H; Klapfer, E

    2003-01-01

    A software implementation of a loss-free counting multichannel analyzer, storing immediately into the multimegabyte memory of a low-cost 486 or Pentium type PC, enables the real-time control of a rabbit system as well as the collection of up to 1000 pairs of simultaneously recorded loss-corrected and non-corrected spectra of 16 k channels each, in a true sequence without time gaps in between, at throughput rates of up to 200 kc/s. Intended for activation analysis of short-lived isomeric transitions, the system renders possible peak to background optimizations and separations of lines with different half-lives without an a priori knowledge of sample composition by summing up appropriate numbers of spectra over appropriate intervals of time. By automatically adapting the noise filtering time to individual pulse intervals, the Preloaded Digital Filter (PLDF) combines low- to medium-rate resolutions comparable to those of high-quality Gaussian amplifiers with throughput rates of up to 100 kc/s, and high-rate reso...

  8. Analysis of three-dimensional thermo-hydraulic phenomena in the reactor core of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, S.; Lee, Y. B.; Jang, W. P.; Ha, K. S.; Jung, H. Y. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    The mismatch between power and flow under the transient condition of LMFBR (Liquid Metal cooled Fast Breeder Reactor) core results in thermal stratification in hot pool. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response, therefore three-dimensional analysis of thermo-hydraulic phenomena is necessary. In this study, the thermo-hydraulic phenomena under normal operating condition and unprotected transient condition of LMFBR is investigated using which is the three-dimensional analysis code, COMMIX-1AR/P. The basic input data is based on the design data of KALIMER-600, which is sodium-cooled fast breeder reactor developed by KAERI. COMMIX-1AR/P code has not a reactivity model and the power and core flowrate must be supplied in the input data. In this study, results of SSC-K calculation is used. The temperature and velocity distributions are calculated and compared with those of SSC-K calculation results. The UTOF(Unprotected Loss Of Flow) accident is calculated using COMMIX-1AR/P and the temperature and velocity distributions in the total reactor core are calculated and the natural circulation mode under this transient condition is investigated.

  9. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  10. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  11. Fluid dynamic analysis of a continuous stirred tank reactor for technical optimization of wastewater digestion.

    Science.gov (United States)

    Hurtado, F J; Kaiser, A S; Zamora, B

    2015-03-15

    Continuous stirred tank reactors (CSTR) are widely used in wastewater treatment plants to reduce the organic matter and microorganism present in sludge by anaerobic digestion. The present study carries out a numerical analysis of the fluid dynamic behaviour of a CSTR in order to optimize the process energetically. The characterization of the sludge flow inside the digester tank, the residence time distribution and the active volume of the reactor under different criteria are determined. The effects of design and power of the mixing system on the active volume of the CSTR are analyzed. The numerical model is solved under non-steady conditions by examining the evolution of the flow during the stop and restart of the mixing system. An intermittent regime of the mixing system, which kept the active volume between 94% and 99%, is achieved. The results obtained can lead to the eventual energy optimization of the mixing system of the CSTR.

  12. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  13. Three dimensional damage mechanics analysis of real life reactor piping components under various loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Durgaprasad, P.V.; Sahu, M.K.; Dutta, B.K. (Reactor Safety Division, Bhabha Atomic Research Centre Trombay, Mumbai (India)), e-mail: pvdp@barc.gov.in

    2009-07-01

    In this work, the Gurson-Tvergaard-Needleman (GTN) damage mechanics model is used for the crack growth analysis of real life reactor piping components. The paper also addresses the challenges involved in such analyses. As a part of component integrity testing, a comprehensive experimental program has been pursued to generate the fracture behavior of reactor piping components. Several real life pipes and elbows with various flaw sizes have been tested under different loading conditions like temperature, pressure, bending etc. In the present work, some of the selected components have been analyzed numerically by using parallel in-house finite element code 'MADAM' with GTN constitutive model. The strength of the micro mechanical models has been demonstrated by comparing the numerical results like load v/s. load-line displacements, J-R curves with the experimental data

  14. Meta-analysis: Its role in psychological methodology

    Directory of Open Access Journals (Sweden)

    Andrej Kastrin

    2008-11-01

    Full Text Available Meta-analysis refers to the statistical analysis of a large collection of independent observations for the purpose of integrating results. The main objectives of this article are to define meta-analysis as a method of data integration, to draw attention to some particularities of its use, and to encourage researchers to use meta-analysis in their work. The benefits of meta-analysis include more effective exploitation of existing data from independent sources and contribution to more powerful domain knowledge. It may also serve as a support tool to generate new research hypothesis. The idea of combining results of independent studies addressing the same research question dates back to sixteenth century. Metaanalysis was reinvented in 1976 by Glass, to refute the conclusion of an eminent colleague, Eysenck, that psychotherapy was essentially ineffective. We review some major historical landmarks of metaanalysis and its statistical background. We present the concept of effect size measure, the problem of heterogeneity and two models which are used to combine individual effect sizes (fixed and random effect model in great details. Two visualization techniques, forest and funnel plot graphics are demonstrated. We developed RMetaWeb, simple and fast web server application to conduct meta-analysis online. RMetaWeb is the first web meta-analysis application and is completely based on R software environment for statistical computing and graphics.

  15. Development of methodology for horizontal axis wind turbine dynamic analysis

    Science.gov (United States)

    Dugundji, J.

    1982-01-01

    Horizontal axis wind turbine dynamics were studied. The following findings are summarized: (1) review of the MOSTAS computer programs for dynamic analysis of horizontal axis wind turbines; (2) review of various analysis methods for rotating systems with periodic coefficients; (3) review of structural dynamics analysis tools for large wind turbine; (4) experiments for yaw characteristics of a rotating rotor; (5) development of a finite element model for rotors; (6) development of simple models for aeroelastics; and (7) development of simple models for stability and response of wind turbines on flexible towers.

  16. Full cost accounting in the analysis of separated waste collection efficiency: A methodological proposal.

    Science.gov (United States)

    D'Onza, Giuseppe; Greco, Giulio; Allegrini, Marco

    2016-02-01

    Recycling implies additional costs for separated municipal solid waste (MSW) collection. The aim of the present study is to propose and implement a management tool - the full cost accounting (FCA) method - to calculate the full collection costs of different types of waste. Our analysis aims for a better understanding of the difficulties of putting FCA into practice in the MSW sector. We propose a FCA methodology that uses standard cost and actual quantities to calculate the collection costs of separate and undifferentiated waste. Our methodology allows cost efficiency analysis and benchmarking, overcoming problems related to firm-specific accounting choices, earnings management policies and purchase policies. Our methodology allows benchmarking and variance analysis that can be used to identify the causes of off-standards performance and guide managers to deploy resources more efficiently. Our methodology can be implemented by companies lacking a sophisticated management accounting system.

  17. Reliability and safety of the K Reactor cooling system: Part 2, Engineering analysis of hydraulic and mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Shoemaker, R.H.

    1960-04-04

    Subsequent to the recent formulation and adoption of safety criteria for reactor cooling systems, there appeared the need for an independent evaluation of the safety and reliability of the K-Reactor cooling system in terms of these criteria. The primary, secondary and last-ditch cooling systems of this reactor involve a strong inter-dependence between electrical and hydraulic components of the water plant. Because of the complexity of inter-relationships between these components, the analysis was divided into two parallel studies which were accomplished during the simmer of 1959. F. D. Robbins has presented his analysis of the electrical power and control system in HW-61887. This report deals with an engineering analysis of the hydraulic and mechanical aspects of the reliability and safety of the K-Reactor Cooling System. The system, as described in this report, is that which existed during the simmer of 1959, prior to modification under Project CG-775 (now Project CG-883).

  18. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary syst