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Sample records for core analysis procedures

  1. A study of the decontamination procedures used for chemical analysis of polar deep ice cores

    Directory of Open Access Journals (Sweden)

    Takayuki Miyake

    2009-11-01

    Full Text Available We investigated the decontamination procedures used on polar deep ice cores before chemical analyses such as measurements of the concentrations of iron species and dust (microparticles. We optimized cutting and melting protocols for decontamination using chemically ultraclean polyethylene bags and simulated ice samples made from ultrapure water. For dust and ion species including acetate, which represented a high level of contamination, we were able to decrease contamination to below several μg l^ for ion concentrations and below 10000 particles ml^ for the dust concentration. These concentration levels of ion species and dust are assumed to be present in the Dome Fuji ice core during interglacial periods. Decontamination of the ice core was achieved by cutting away approximately 3 mm of the outside of a sample and by melting away approximately 30% of a sample's weight. Furthermore, we also report the preparation protocols for chemical analyses of the 2nd Dome Fuji ice core, including measurements of ion and dust concentrations, pH, electric conductivity (EC, and stable isotope ratios of water (δD and δO, based on the results of the investigation of the decontamination procedures.

  2. Core calculational techniques and procedures

    International Nuclear Information System (INIS)

    Romano, J.J.

    1977-10-01

    Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range

  3. Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook

    2005-01-01

    A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem

  4. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  5. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  6. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

    Cook, B.A.; Carlson, E.R.

    1985-01-01

    One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

  7. Stereotactic large-core needle breast biopsy: analysis of pain and discomfort related to the biopsy procedure

    International Nuclear Information System (INIS)

    Hemmer, Judith M.; Heesewijk, Hans P.M. van; Kelder, Johannes C.

    2008-01-01

    The purpose of this study was to determine the significance of variables such as duration of the procedure, type of breast tissue, number of passes, depth of the biopsies, underlying pathology, the operator performing the procedure, and their effect on women's perception of pain and discomfort during stereotactic large-core needle breast biopsy. One hundred and fifty consecutive patients with a non-palpable suspicious mammographic lesions were included. Between three and nine 14-gauge breast passes were taken using a prone stereotactic table. Following the biopsy procedure, patients were asked to complete a questionnaire. There was no discomfort in lying on the prone table. There is no relation between type of breast lesion and pain, underlying pathology and pain and performing operator and pain. The type of breast tissue is correlated with pain experienced from biopsy (P = 0.0001). We found out that patients with dense breast tissue complain of more pain from biopsy than patients with more involution of breast tissue. The depth of the biopsy correlates with pain from biopsy (P = 0.0028). Deep lesions are more painful than superficial ones. There is a correlation between the number of passes and pain in the neck (P = 0.0188) and shoulder (P = 0.0366). The duration of the procedure is correlated with pain experienced in the neck (P = 0.0116) but not with pain experienced from biopsy. (orig.)

  8. Acceptance test procedure for core sample trucks

    International Nuclear Information System (INIS)

    Smalley, J.L.

    1995-01-01

    The purpose of this Acceptance Test Procedure is to provide instruction and documentation for acceptance testing of the rotary mode core sample trucks, HO-68K-4600 and HO-68K-4647. The rotary mode core sample trucks were based upon the design of the second core sample truck (HO-68K-4345) which was constructed to implement rotary mode sampling of the waste tanks at Hanford. Acceptance testing of the rotary mode core sample trucks will verify that the design requirements have been met. All testing will be non-radioactive and stand-in materials shall be used to simulate waste tank conditions. Compressed air will be substituted for nitrogen during the majority of testing, with nitrogen being used only for flow characterization

  9. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  10. LMFBR core design analysis

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Licensing procedures and safety criteria for core conversion in Japan

    International Nuclear Information System (INIS)

    Kanda, K.; Nakagome, Y.; Hayashi, M.

    1983-01-01

    Procedures relating to the construction and operation of reactor facilities are discussed. Specifically, the Safety Analysis Report on the Kyoto University Critical Assembly (KUCA) core conversion (93% to 45% enrichment) is noted. The results of critical experiments in the KUCA and of burnup tests in the Oak Ridge Research (ORR) Reactor will be used in the final determination of the feasibility of the conversion of the Kyoto University High Flux Reactor (KUHFR) to the use of 45% enrichment

  12. Licensing procedures and safety criteria for core conversion in Japan

    International Nuclear Information System (INIS)

    Kanda, K.; Nakagome, Y.; Hayashi, M.

    1983-01-01

    In Japan, the establishment and operation of nuclear installations are governed mainly by the Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors. This law lays down the regulations and conditions for licensing of the various installations involved in the nuclear fuel cycle, namely licensing of installations for refining, fabricating and reprocessing; and reactors, as well as licensing of the use of nuclear fuels in research facilities. Although procedures for the installations listed above vary depending on the installation concerned, only those relating to construction and operation of reactor facilities will be analysed in this study, as the conditions and principles applying to licensing and control of other installations are, to a large extent, similar to those concerning reactor facilities. The second part of this presentation describes the safety review of the KUCA reactor core conversion form HEU to MEU. For the safety review of the core conversion, the Committee on Examination of Reactor Safety of Japanese Government examined mainly the the nuclear characteristics and the integrity of aluminide fuel plates, which was very severe because we had no experience to use aluminide fuel plates in Japan. The integrity of fuel plates and the results of the worst accident analysis for the MEU core are shown with the comparison between the HEU and MEU cores. The significant difference was not observed between them. All the regulatory procedures were completed in September 1980. Fabrication of MEU fuel elements for the KUCA experiments by CERCA in France was started in September 1980, and will be completed in March 1981. The critical experiments in the KUCA with MEU fuel will be started on a single-core in May 1981 as a first step. Those on a coupled-core will follow

  13. Core on-line monitoring and computerized procedures systems

    International Nuclear Information System (INIS)

    Gangloff, W.C.

    1986-01-01

    The availability of operating nuclear power plants has been affected significantly by the difficulty people have in coping with the complexity of the plants and the operating procedures. Two ways to use modern computer technology to ease the burden of coping are discussed in this paper, an on-line core monitoring system with predictive capability and a computerized procedures system using live plant data. These systems reduce human errors by presenting information rather than simply data, using the computer to manipulate the data, but leaving the decisions to the plant operator

  14. Probabilistic safety analysis procedures guide

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Buslik, A.J.

    1984-01-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of tissues affecting reactor safety. This guide addresses the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant and from loss of offsite electric power. The scope includes analyses of problem-solving (cognitive) human errors, a determination of importance of the various core damage accident sequences, and an explicit treatment and display of uncertainties for the key accident sequences. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance) and the risk associated with external accident initiators, as consensus is developed regarding suitable methodologies in these areas. This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are essential for regulatory decision making. Methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study

  15. Gas Hydrate Investigations Using Pressure Core Analysis: Current Practice

    Science.gov (United States)

    Schultheiss, P.; Holland, M.; Roberts, J.; Druce, M.

    2006-12-01

    Recently there have been a number of major gas hydrate expeditions, both academic and commercially oriented, that have benefited from advances in the practice of pressure coring and pressure core analysis, especially using the HYACINTH pressure coring systems. We report on the now mature process of pressure core acquisition, pressure core handling and pressure core analysis and the results from the analysis of pressure cores, which have revealed important in situ properties along with some remarkable views of gas hydrate morphologies. Pressure coring success rates have improved as the tools have been modified and adapted for use on different drilling platforms. To ensure that pressure cores remain within the hydrate stability zone, tool deployment, recovery and on-deck handling procedures now mitigate against unwanted temperature rises. Core analysis has been integrated into the core transfer protocol and automated nondestructive measurements, including P-wave velocity, gamma density, and X-ray imaging, are routinely made on cores. Pressure cores can be subjected to controlled depressurization experiments while nondestructive measurements are being made, or cores can be stored at in situ conditions for further analysis and subsampling.

  16. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  17. CT-guided core needle biopsy of mediastinal nodes through a transpulmonary approach: retrospective analysis of the procedures conducted over six years.

    Science.gov (United States)

    Yin, Zhongyuan; Liang, Zhiwen; Li, Pengcheng; Wang, Qiong

    2017-08-01

    To retrospectively evaluate the diagnostic performance and complications of a CT-guided core needle cutting biopsy of mediastinal nodes through a transpulmonary approach. From January 2009 to December 2014, we used a coaxial positioning system and an 18G cutting-type biopsy device to perform CT-guided percutaneous transpulmonary needle biopsies of mediastinal nodes for 127 patients. The diagnostic performance, complication rate, influencing factors, distribution of mediastinal nodes and pathological diagnoses were investigated. Among 127 patients, pathologic analyses showed that all of the biopsies were technically successful. The sensitivity, specificity, positive predictive value, and negative predictive value were all 100%. As for complications, the ratios for pneumothorax and hemoptysis were 33.9% and 4.7%, respectively. Multivariate analyses revealed that the distance from the pleura to the target lesion (P = 0.008) and the numbers of visceral pleura injuries (P = 0.006) were the two most significant risk factors for pneumothorax, and that the distance from the pleura to the target lesion (P = 0.004) was the most significant risk factor for hemoptysis. CT-guided core needle cutting biopsy of mediastinal nodes through a transpulmonary approach is a safe and efficient diagnostic method. • CT-guided core needle biopsy is an accurate technique for diagnosing mediastinal nodes. • The rates of complications are similar to those for pulmonary lesion biopsy. • Pneumothorax risk factors include distance from pleura to target lesion and number of visceral pleura. • Distance from pleura to target lesion is the risk factor for hemoptysis. • CT-guided core needle biopsy is an important diagnostic method for mediastinal nodes.

  18. CT-guided core needle biopsy of mediastinal nodes through a transpulmonary approach: retrospective analysis of the procedures conducted over six years

    Energy Technology Data Exchange (ETDEWEB)

    Yin, Zhongyuan; Liang, Zhiwen; Li, Pengcheng; Wang, Qiong [Huazhong University of Science and Technology, Cancer Center, Union Hospital, Tongji Medical College, Wuhan (China)

    2017-08-15

    To retrospectively evaluate the diagnostic performance and complications of a CT-guided core needle cutting biopsy of mediastinal nodes through a transpulmonary approach. From January 2009 to December 2014, we used a coaxial positioning system and an 18G cutting-type biopsy device to perform CT-guided percutaneous transpulmonary needle biopsies of mediastinal nodes for 127 patients. The diagnostic performance, complication rate, influencing factors, distribution of mediastinal nodes and pathological diagnoses were investigated. Among 127 patients, pathologic analyses showed that all of the biopsies were technically successful. The sensitivity, specificity, positive predictive value, and negative predictive value were all 100%. As for complications, the ratios for pneumothorax and hemoptysis were 33.9% and 4.7%, respectively. Multivariate analyses revealed that the distance from the pleura to the target lesion (P = 0.008) and the numbers of visceral pleura injuries (P = 0.006) were the two most significant risk factors for pneumothorax, and that the distance from the pleura to the target lesion (P = 0.004) was the most significant risk factor for hemoptysis. CT-guided core needle cutting biopsy of mediastinal nodes through a transpulmonary approach is a safe and efficient diagnostic method. (orig.)

  19. Development of a Web-based CANDU Core Management Procedure Automation System

    International Nuclear Information System (INIS)

    Lee, Sanghoon; Kim, Eunggon; Park, Daeyou; Yeom, Choongsub; Suh, Hyungbum; Kim, Sungmin

    2006-01-01

    CANDU reactor core needs efficient core management to increase safety, stability, high performance as well as to decrease operational cost. The most characteristic feature of CANDU is so called 'on-power refueling' i.e., there is no shutdown during refueling in opposition to that of PWR. Although this on-power refueling increases the efficiency of the plant, it requires heavy operational task and difficulties in real time operation such as regulating power distribution, burnup distribution, LZC statistics, the position of control devices and so on. To enhance the CANDU core management, there are several approaches to help operator and reduce difficulties, one of them is the COMOS (CANDU Core On-line Monitoring System). It has developed as an online core surveillance system based on the standard incre instrumentation and the numerical analysis codes such as RFSP (Reactor Fueling Simulation Program). As the procedure is getting more complex and the number of programs is increased, it is required that integrated and cooperative system. So, KHNP and IAE have been developing a new web-based system which can support effective and accurate reactor operational environment called COMPAS that means CANDU cOre Management Procedure Automation System. To ensure development of successful system, several steps of identifying requirements have been performed and Software Requirement Specification (SRS) document was developed. In this paper we emphasis on the how to keep consistency between the requirements and system products by applying requirement traceability methodology

  20. Biomass Compositional Analysis Laboratory Procedures | Bioenergy | NREL

    Science.gov (United States)

    Biomass Compositional Analysis Laboratory Procedures Biomass Compositional Analysis Laboratory Procedures NREL develops laboratory analytical procedures (LAPs) for standard biomass analysis. These procedures help scientists and analysts understand more about the chemical composition of raw biomass

  1. Development of a web-based CANDU core management procedures automation system

    International Nuclear Information System (INIS)

    Lee, S.; Park, D.; Yeom, C.; Suh, H.

    2007-01-01

    Introduce CANDU core management procedures automation system (COMPAS) - A web-based application which semi-automates several CANDU core management tasks. It provides various functionalities including selection and evaluation of refueling channel, detector calibration, coolant flow estimation and thermal power calculation through automated interfacing with analysis codes (RFSP, NUCIRC, etc.) and plant data. It also utilizes brand new .NET computing technology such as ASP.NET, smart client, web services and so on. Since almost all functions are abstracted from the previous experiences of the current working members of the Wolsong Nuclear Power Plant (NPP), it will lead to an efficient and safe operation of CANDU plants. (author)

  2. Development of a web-based CANDU core management procedures automation system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.; Park, D.; Yeom, C. [Inst. for Advanced Engineering (IAE), Yongin (Korea, Republic of); Suh, H. [Korea Hydro and Nuclear Power (KHNP), Wolsong (Korea, Republic of)

    2007-07-01

    Introduce CANDU core management procedures automation system (COMPAS) - A web-based application which semi-automates several CANDU core management tasks. It provides various functionalities including selection and evaluation of refueling channel, detector calibration, coolant flow estimation and thermal power calculation through automated interfacing with analysis codes (RFSP, NUCIRC, etc.) and plant data. It also utilizes brand new .NET computing technology such as ASP.NET, smart client, web services and so on. Since almost all functions are abstracted from the previous experiences of the current working members of the Wolsong Nuclear Power Plant (NPP), it will lead to an efficient and safe operation of CANDU plants. (author)

  3. Core analysis: new features and applications

    International Nuclear Information System (INIS)

    Edenius, M.; Kurcyusz, E.; Molina, D.; Wiksell, G.

    1995-01-01

    Today, core analysis may be performed with sophisticated software capable of both steady state and transient analysis using a common methodology for BWRs and PWRs. General trends in core analysis software development are: improved accuracy, automated engineering functions; three-dimensional transient capability; graphical user interfaces. As a demonstration of such software, new features of Studsvik-CMS (Core management system) and examples of applications are discussed in this article. 2 figs., 8 refs

  4. Safety analysis procedures for PHWR

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

    2004-03-01

    The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

  5. Procedures monitoring and MAAP analysis

    International Nuclear Information System (INIS)

    May, R.S.

    1991-01-01

    Numerous studies of severe accidents in light water reactors have shown that operator response can play a crucial role in the predicted outcomes of dominant accident scenarios. MAAP provides the capability to specify certain operator actions as input data. However, making reasonable assumptions about the nature and timing of operator response requires substantial knowledge about plant practices and procedures and what they imply for the event being analyzed. The appearance of knowledge based software technology in the mid-1980s provided a natural format for representing and maintaining procedures as IF-THEN rules. The boiling water reactor (BWR) Emergency Operating Procedures Tracking System (EOPTS) was composed of a rule base of procedures and a dedicated inference engine (problem-solver). Based on the general approach and experience of EOPTS, the authors have developed a prototype procedures monitoring system that reads MAAP transient output files and evaluate the EOP messages and instructions that would be implied during each transient time interval. The prototype system was built using the NEXPERT OBJECT expert system development system, running on a 386-class personal computer with 4 MB of memory. The limited scope prototype includes a reduced set of BWR6 EOPs procedures evaluation on a coarse time interval, a simple text-based user interface, and a summary-report generator. The prototype, which is limited to batch-mode analysis of MAAP output, is intended to demonstrate the concept and aid in the design of a production system, which will involve a direct link to MAAP and interactive capabilities

  6. An analysis of the uniform core experiment

    Energy Technology Data Exchange (ETDEWEB)

    Waterson, R H

    1973-10-15

    This report describes an analysis of the Uniform Core of HITREX using the WIMS E codes, and presents the results of theory/experiment comparisons. The overall picture is one of good agreement for core reaction rate distributions, but theory umderestimating k{sub eff} by about 1.5% {delta}k/k.

  7. Statistical evaluations of current sampling procedures and incomplete core recovery

    International Nuclear Information System (INIS)

    Heasler, P.G.; Jensen, L.

    1994-03-01

    This document develops two formulas that describe the effects of incomplete recovery on core sampling results for the Hanford waste tanks. The formulas evaluate incomplete core recovery from a worst-case (i.e.,biased) and best-case (i.e., unbiased) perspective. A core sampler is unbiased if the sample material recovered is a random sample of the material in the tank, while any sampler that preferentially recovers a particular type of waste over others is a biased sampler. There is strong evidence to indicate that the push-mode sampler presently used at the Hanford site is a biased one. The formulas presented here show the effects of incomplete core recovery on the accuracy of composition measurements, as functions of the vertical variability in the waste. These equations are evaluated using vertical variability estimates from previously sampled tanks (B110, U110, C109). Assuming that the values of vertical variability used in this study adequately describes the Hanford tank farm, one can use the formulas to compute the effect of incomplete recovery on the accuracy of an average constituent estimate. To determine acceptable recovery limits, we have assumed that the relative error of such an estimate should be no more than 20%

  8. Updated procedures for using drill cores and cuttings at the Lithologic Core Storage Library, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Hodges, Mary K.V.; Davis, Linda C.; Bartholomay, Roy C.

    2018-01-30

    In 1990, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy Idaho Operations Office, established the Lithologic Core Storage Library at the Idaho National Laboratory (INL). The facility was established to consolidate, catalog, and permanently store nonradioactive drill cores and cuttings from subsurface investigations conducted at the INL, and to provide a location for researchers to examine, sample, and test these materials.The facility is open by appointment to researchers for examination, sampling, and testing of cores and cuttings. This report describes the facility and cores and cuttings stored at the facility. Descriptions of cores and cuttings include the corehole names, corehole locations, and depth intervals available.Most cores and cuttings stored at the facility were drilled at or near the INL, on the eastern Snake River Plain; however, two cores drilled on the western Snake River Plain are stored for comparative studies. Basalt, rhyolite, sedimentary interbeds, and surficial sediments compose most cores and cuttings, most of which are continuous from land surface to their total depth. The deepest continuously drilled core stored at the facility was drilled to 5,000 feet below land surface. This report describes procedures and researchers' responsibilities for access to the facility and for examination, sampling, and return of materials.

  9. A procedure for searching the equilibrium core of a research reactor

    International Nuclear Information System (INIS)

    Bakri Arbie; Liem Peng Hong; Prayoto

    1996-01-01

    A procedure for searching the equilibrium core of a research reactor has been proposed. The searching procedure is based on the relaxation method and has been implemented in Batan-EQUIL-2D in-core fuel management code. The few-group neutron diffusion theory in 2-D, X-Y, and R-Z reactor geometries are adopted as the framework of the code. The successful applicability of the procedure for obtaining the new RSG-GAS (MPR-30) silicide equilibrium core was shown. (author)

  10. Operability test procedure for the Rotary Mode Core Sampling System Exhausters 3 and 4

    International Nuclear Information System (INIS)

    WSaldo, E.J.

    1995-01-01

    This document provides a procedure for performing operability testing of the Rotary Mode Core Sampling System Exhausters 3 ampersand 4. Upon completion of testing activities an operability testing report will be issued

  11. Core Competencies or a Competent Core? A Scoping Review and Realist Synthesis of Invasive Bedside Procedural Skills Training in Internal Medicine.

    Science.gov (United States)

    Brydges, Ryan; Stroud, Lynfa; Wong, Brian M; Holmboe, Eric S; Imrie, Kevin; Hatala, Rose

    2017-11-01

    Invasive bedside procedures are core competencies for internal medicine, yet no formal training guidelines exist. The authors conducted a scoping review and realist synthesis to characterize current training for lumbar puncture, arthrocentesis, paracentesis, thoracentesis, and central venous catheterization. They aimed to collate how educators justify using specific interventions, establish which interventions have the best evidence, and offer directions for future research and training. The authors systematically searched Medline, Embase, the Cochrane Library, and ERIC through April 2015. Studies were screened in three phases; all reviews were performed independently and in duplicate. The authors extracted information on learner and patient demographics, study design and methodological quality, and details of training interventions and measured outcomes. A three-step realist synthesis was performed to synthesize findings on each study's context, mechanism, and outcome, and to identify a foundational training model. From an initial 6,671 studies, 149 studies were further reduced to 67 (45%) reporting sufficient information for realist synthesis. Analysis yielded four types of procedural skills training interventions. There was relative consistency across contexts and significant differences in mechanisms and outcomes across the four intervention types. The medical procedural service was identified as an adaptable foundational training model. The observed heterogeneity in procedural skills training implies that programs are not consistently developing residents who are competent in core procedures. The findings suggest that researchers in education and quality improvement will need to collaborate to design training that develops a "competent core" of proceduralists using simulation and clinical rotations.

  12. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  13. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

    Waldman, R.M.; Vertullo, A.C.

    1987-01-01

    The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

  14. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Lee, T.T.; Schwartz, A.A.; Koopman, D.C.A.

    1980-05-01

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  15. Core instrumentation and pre-operational procedures for core conversion HEU to LEU

    International Nuclear Information System (INIS)

    1984-02-01

    This report is intended for the reactor operator, to be used as a manual or checklist for general guidance on pre-startup activities that need to be addressed in preparation for conversion to Low Enriched Fuel (LEU). All nuclear, thermodynamic and safety calculations should have been performed prior to this stage of the core conversion process. During these calculations and certainly before ordering the new LEU fuel elements the reactor operator needs to very carefully consider additional important factors concerning the new fuel: fuel reliability, reliability of fuel fabricator, reprocessing contract or fuel element storage and disposal, economics of the new fuel cycle. At this stage, too, a preoperational experimental programme has to be developed and presented to the regulatory authorities for approval. This experimental programme could lead to additional requirements on: in-core instrumentation, out-of-core instrumentation or additional experimental devices. Detailed instructions on specific tests and measurements are not provided in this report since much information on the subject is available in the open literature

  16. Statistical hot spot analysis of reactor cores

    International Nuclear Information System (INIS)

    Schaefer, H.

    1974-05-01

    This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core

  17. Analysis of emergency operating procedures effectiveness for core damage prevention using computer code RELAP for nuclear power plants with VVER-1000/B-320 in reference to primary to secondary circuit leak with external power loss and BRU-A stuck open failure

    International Nuclear Information System (INIS)

    Arkhangelski, L.; Sheveliov, D. V.

    1999-01-01

    This report presents analysis of development emergency operating procedures effectiveness for possible accident on nuclear power plant with WWER-1000 reactor type. Accident initiating event is the primary to secondary circuit leak caused by steam generator primary cover lift-up. In according to conservative assumptions the following additional failures were considered: dump valve BRU-A stuck open failure; loss of external power. The results of this work are represented as a comparative analysis of two possible ways of accident evolution: according to functioning automatic safety systems responses; according to accident management based on development emergency operating procedures with operator intervention. Developed emergency operating procedures assure the following significant goals to mitigate accident sequences: optimal use of ECCS water inventory; severe core damage prevention; mitigation of environment radioactive contamination. (authors)

  18. Preliminary analysis of the proposed BN-600 benchmark core

    International Nuclear Information System (INIS)

    John, T.M.

    2000-01-01

    The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

  19. Proposed Core Competencies and Empirical Validation Procedure in Competency Modeling: Confirmation and Classification.

    Science.gov (United States)

    Baczyńska, Anna K; Rowiński, Tomasz; Cybis, Natalia

    2016-01-01

    Competency models provide insight into key skills which are common to many positions in an organization. Moreover, there is a range of competencies that is used by many companies. Researchers have developed core competency terminology to underline their cross-organizational value. The article presents a theoretical model of core competencies consisting of two main higher-order competencies called performance and entrepreneurship. Each of them consists of three elements: the performance competency includes cooperation, organization of work and goal orientation, while entrepreneurship includes innovativeness, calculated risk-taking and pro-activeness. However, there is lack of empirical validation of competency concepts in organizations and this would seem crucial for obtaining reliable results from organizational research. We propose a two-step empirical validation procedure: (1) confirmation factor analysis, and (2) classification of employees. The sample consisted of 636 respondents (M = 44.5; SD = 15.1). Participants were administered a questionnaire developed for the study purpose. The reliability, measured by Cronbach's alpha, ranged from 0.60 to 0.83 for six scales. Next, we tested the model using a confirmatory factor analysis. The two separate, single models of performance and entrepreneurial orientations fit quite well to the data, while a complex model based on the two single concepts needs further research. In the classification of employees based on the two higher order competencies we obtained four main groups of employees. Their profiles relate to those found in the literature, including so-called niche finders and top performers. Some proposal for organizations is discussed.

  20. Analysis of core damage frequency: Surry, Unit 1 internal events

    International Nuclear Information System (INIS)

    Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

    1990-04-01

    This document contains the accident sequence analysis of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed and described here is an extensive of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments form numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.05-E-5 per year, with a 95% upper bound of 1.34E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency. 49 refs., 52 figs., 70 tabs

  1. Human Reliability Analysis For Computerized Procedures

    International Nuclear Information System (INIS)

    Boring, Ronald L.; Gertman, David I.; Le Blanc, Katya

    2011-01-01

    This paper provides a characterization of human reliability analysis (HRA) issues for computerized procedures in nuclear power plant control rooms. It is beyond the scope of this paper to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper provides a review of HRA as applied to traditional paper-based procedures, followed by a discussion of what specific factors should additionally be considered in HRAs for computerized procedures. Performance shaping factors and failure modes unique to computerized procedures are highlighted. Since there is no definitive guide to HRA for paper-based procedures, this paper also serves to clarify the existing guidance on paper-based procedures before delving into the unique aspects of computerized procedures.

  2. Development of UCMS for Analysis of Designed and Measured Core Power Distribution

    International Nuclear Information System (INIS)

    Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

    2009-01-01

    In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

  3. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    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

  4. A retrospective analysis of ultrasound-guided large core needle ...

    African Journals Online (AJOL)

    2016-07-27

    Jul 27, 2016 ... The different types of non-surgical breast biopsy procedures include: fine needle aspiration biopsy. (FNAB), core needle ... needle biopsies of breast lesions at a regional public hospital in ..... NCR_2009_FINAL.pdf. 2. Parikh J ...

  5. Simplified procedures for fast reactor fuel cycle and sensitivity analysis

    International Nuclear Information System (INIS)

    Badruzzaman, A.

    1979-01-01

    The Continuous Slowing Down-Integral Transport Theory has been extended to perform criticality calculations in a Fast Reactor Core-blanket system achieving excellent prediction of the spectrum and the eigenvalue. The integral transport parameters did not need recalculation with source iteration and were found to be relatively constant with exposure. Fuel cycle parameters were accurately predicted when these were not varied, thus reducing a principal potential penalty of the Intergal Transport approach where considerable effort may be required to calculate transport parameters in more complicated geometries. The small variation of the spectrum in the central core region, and its weak dependence on exposure for both this region, the core blanket interface and blanket region led to the extension and development of inexpensive simplified procedures to complement exact methods. These procedures gave accurate predictions of the key fuel cycle parameters such as cost and their sensitivity to variation in spectrum-averaged and multigroup cross sections. They also predicted the implications of design variation on these parameters very well. The accuracy of these procedures and their use in analyzing a wide variety of sensitivities demonstrate the potential utility of survey calculations in Fast Reactor analysis and fuel management

  6. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  7. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  8. Multi-Core Processor Memory Contention Benchmark Analysis Case Study

    Science.gov (United States)

    Simon, Tyler; McGalliard, James

    2009-01-01

    Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

  9. Reactivity analysis of core distortion effects in the FFTF

    International Nuclear Information System (INIS)

    Knutson, B.J.

    1982-01-01

    An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analysis method

  10. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2015-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author).

  11. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2014-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author)

  12. Bio-Oil Analysis Laboratory Procedures | Bioenergy | NREL

    Science.gov (United States)

    Bio-Oil Analysis Laboratory Procedures Bio-Oil Analysis Laboratory Procedures NREL develops laboratory analytical procedures (LAPs) for the analysis of raw and upgraded pyrolysis bio-oils. These standard procedures have been validated and allow for reliable bio-oil analysis. Procedures Determination

  13. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  14. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

    Lin, J.H.; Huang, R.L.; Sawyer, C.D.

    1993-01-01

    This paper reports on a thermal hydraulic analysis of the SIMPLIFIED BOILING WATER REACTOR (SBWR) during startup. The potential instability during a SBWR startup has drawn the attention of designers, researchers, and engineers. It has not been a concern for a Boiling Water Reactor (BWR) with forced recirculation; however, for SBWR with natural circulation the concern exists. The concern is about the possibility of a geysering mode oscillation during SBWR startup from a cold temperature and a low system pressure with a low natural circulation flow rate. A thermal hydraulic analysis of the SBWR is performed in simulation of the startup using the TRACG computer code. The temperature, pressure, and reactor power profiles of SBWR during the startup are presented. The results are compared with the data of a natural circulation boiling water reactor, the DODEWAARD plant, in which no instabilities have been observed during many startups. It is shown that a SBWR startup which follows proper procedures, geysering and other modes of oscillations can be avoided

  15. Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, T; Ito, H [Geological Survey of Japan, Tsukuba (Japan)

    1996-05-01

    This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

  16. Procedures for use of, and drill cores and cuttings available for study at, the Lithologic Core Storage Library, Idaho National Engineering Laboratory, Idaho

    International Nuclear Information System (INIS)

    Davis, L.C.; Hannula, S.R.; Bowers, B.

    1997-03-01

    In 1990, the US Geological Survey, in cooperation with the US Department of Energy, Idaho Operations Office, established the Lithologic Core Storage Library at the Idaho National Engineering Laboratory (INEL). The facility was established to consolidate, catalog, and permanently store nonradioactive drill cores and cuttings from investigations of the subsurface conducted at the INEL, and to provide a location for researchers to examine, sample, and test these materials. The facility is open by appointment to researchers for examination, sampling, and testing of cores and cuttings. This report describes the facility and cores and cuttings stored at the facility. Descriptions of cores and cuttings include the well names, well locations, and depth intervals available. Most cores and cuttings stored at the facility were drilled at or near the INEL, on the eastern Snake River Plain; however, two cores drilled on the western Snake River Plain are stored for comparative studies. Basalt, rhyolite, sedimentary interbeds, and surficial sediments compose the majority of cores and cuttings, most of which are continuous from land surface to their total depth. The deepest core stored at the facility was drilled to 5,000 feet below land surface. This report describes procedures and researchers' responsibilities for access to the facility, and examination, sampling, and return of materials

  17. Safeguards Network Analysis Procedure (SNAP): overview

    International Nuclear Information System (INIS)

    Chapman, L.D; Engi, D.

    1979-08-01

    Nuclear safeguards systems provide physical protection and control of nuclear materials. The Safeguards Network Analysis Procedure (SNAP) provides a convenient and standard analysis methodology for the evaluation of physical protection system effectiveness. This is achieved through a standard set of symbols which characterize the various elements of safeguards systems and an analysis program to execute simulation models built using the SNAP symbology. The outputs provided by the SNAP simulation program supplements the safeguards analyst's evaluative capabilities and supports the evaluation of existing sites as well as alternative design possibilities. This paper describes the SNAP modeling technique and provides an example illustrating its use

  18. Statistical analysis of dynamic parameters of the core

    International Nuclear Information System (INIS)

    Ionov, V.S.

    2007-01-01

    The transients of various types were investigated for the cores of zero power critical facilities in RRC KI and NPP. Dynamic parameters of neutron transients were explored by tool statistical analysis. Its have sufficient duration, few channels for currents of chambers and reactivity and also some channels for technological parameters. On these values the inverse period. reactivity, lifetime of neutrons, reactivity coefficients and some effects of a reactivity are determinate, and on the values were restored values of measured dynamic parameters as result of the analysis. The mathematical means of statistical analysis were used: approximation(A), filtration (F), rejection (R), estimation of parameters of descriptive statistic (DSP), correlation performances (kk), regression analysis(KP), the prognosis (P), statistician criteria (SC). The calculation procedures were realized by computer language MATLAB. The reasons of methodical and statistical errors are submitted: inadequacy of model operation, precision neutron-physical parameters, features of registered processes, used mathematical model in reactivity meters, technique of processing for registered data etc. Examples of results of statistical analysis. Problems of validity of the methods used for definition and certification of values of statistical parameters and dynamic characteristics are considered (Authors)

  19. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  20. Event course analysis of core disruptive accidents

    International Nuclear Information System (INIS)

    Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

    1995-01-01

    The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

  1. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2004-01-01

    Ice cores are the most direct, continuous, and high resolution archive for Late Quaternary paleoclimate reconstruction. Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate migration strategies for New Zealand. (author). 23 refs., 15 figs., 1 tab

  2. Ultrasonography-guided core needle biopsy for the thyroid nodule: does the procedure hold any benefit for the diagnosis when fine-needle aspiration cytology analysis shows inconclusive results?

    Science.gov (United States)

    Hahn, S Y; Han, B-K; Ko, E Y; Ko, E S

    2013-01-01

    Objective: We evaluated the diagnostic role of ultrasonography-guided core needle biopsy (CNB) according to ultrasonography features of thyroid nodules that had inconclusive ultrasonography-guided fine-needle aspiration (FNA) results. Methods: A total of 88 thyroid nodules in 88 patients who underwent ultrasonography-guided CNB because of previous inconclusive FNA results were evaluated. The patients were classified into three groups based on ultrasonography findings: Group A, which was suspicious for papillary thyroid carcinoma (PTC); Group B, which was suspicious for follicular (Hurthle cell) neoplasm; and Group C, which was suspicious for lymphoma. The final diagnoses of the thyroid nodules were determined by surgical confirmation or follow-up after ultrasonography-guided CNB. Results: Of the 88 nodules, the malignant rate was 49.1% in Group A, 12.0% in Group B and 90.0% in Group C. The rates of conclusive ultrasonography-guided CNB results after previous incomplete ultrasonography-guided FNA results were 96.2% in Group A, 64.0% in Group B and 90.0% in Group C (p=0.001). 12 cases with inconclusive ultrasonography-guided CNB results were finally diagnosed as 8 benign lesions, 3 PTCs and 1 lymphoma. The number of previous ultrasonography-guided FNA biopsies was not significantly different between the conclusive and the inconclusive result groups of ultrasonography-guided CNB (p=0.205). Conclusion: Ultrasonography-guided CNB has benefit for the diagnosis of thyroid nodules with inconclusive ultrasonography-guided FNA results. However, it is still not helpful for the differential diagnosis in 36% of nodules that are suspicious for follicular neoplasm seen on ultrasonography. Advances in knowledge: This study shows the diagnostic contribution of ultrasonography-guided CNB as an alternative to repeat ultrasonography-guided FNA or surgery. PMID:23564885

  3. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  4. s-core network decomposition: A generalization of k-core analysis to weighted networks

    Science.gov (United States)

    Eidsaa, Marius; Almaas, Eivind

    2013-12-01

    A broad range of systems spanning biology, technology, and social phenomena may be represented and analyzed as complex networks. Recent studies of such networks using k-core decomposition have uncovered groups of nodes that play important roles. Here, we present s-core analysis, a generalization of k-core (or k-shell) analysis to complex networks where the links have different strengths or weights. We demonstrate the s-core decomposition approach on two random networks (ER and configuration model with scale-free degree distribution) where the link weights are (i) random, (ii) correlated, and (iii) anticorrelated with the node degrees. Finally, we apply the s-core decomposition approach to the protein-interaction network of the yeast Saccharomyces cerevisiae in the context of two gene-expression experiments: oxidative stress in response to cumene hydroperoxide (CHP), and fermentation stress response (FSR). We find that the innermost s-cores are (i) different from innermost k-cores, (ii) different for the two stress conditions CHP and FSR, and (iii) enriched with proteins whose biological functions give insight into how yeast manages these specific stresses.

  5. Building America House Performance Analysis Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Hendron, R.; Farrar-Nagy, S.; Anderson, R.; Judkoff, R.

    2001-10-29

    As the Building America Program has grown to include a large and diverse cross section of the home building industry, accurate and consistent analysis techniques have become more important to help all program partners as they perform design tradeoffs and calculate energy savings for prototype houses built as part of the program. This document illustrates some of the analysis concepts proven effective and reliable for analyzing the transient energy usage of advanced energy systems as well as entire houses. The analysis procedure described here provides a starting point for calculating energy savings of a prototype house relative to two base cases: builder standard practice and regional standard practice. Also provides building simulation analysis to calculate annual energy savings based on side-by-side short-term field testing of a prototype house.

  6. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 45 refs., 16 figs., 2 tabs.

  7. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 27 refs., 18 figs., 2 tabs

  8. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2012-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 28 refs., 20 figs., 1 tab.

  9. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2008-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 27 refs., 18 figs., 2 tabs

  10. Analysis and study on core power capability with margin method

    International Nuclear Information System (INIS)

    Liu Tongxian; Wu Lei; Yu Yingrui; Zhou Jinman

    2015-01-01

    Core power capability analysis focuses on the power distribution control of reactor within the given mode of operation, for the purpose of defining the allowed normal operating space so that Condition Ⅰ maneuvering flexibility is maintained and Condition Ⅱ occurrences are adequately protected by the reactor protection system. For the traditional core power capability analysis methods, such as synthesis method or advanced three dimension method, usually calculate the key safety parameters of the power distribution, and then verify that these parameters meet the design criteria. For PWR with on-line power distribution monitoring system, core power capability analysis calculates the most power level which just meets the design criteria. On the base of 3D FAC method of Westinghouse, the calculation model of core power capability analysis with margin method is introduced to provide reference for engineers. The core power capability analysis of specific burnup of Sanmen NPP is performed with the margin method. The results demonstrate the rationality of the margin method. The calculation model of the margin method not only helps engineers to master the core power capability analysis for AP1000, but also provides reference for engineers for core power capability analysis of other PWR with on-line power distribution monitoring system. (authors)

  11. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  12. Buckling analysis of laminated sandwich beam with soft core

    Directory of Open Access Journals (Sweden)

    Anupam Chakrabarti

    Full Text Available Stability analysis of laminated soft core sandwich beam has been studied by a C0 FE model developed by the authors based on higher order zigzag theory (HOZT. The in-plane displacement variation is considered to be cubic for the face sheets and the core, while transverse displacement is quadratic within the core and constant in the faces beyond the core. The proposed model satisfies the condition of stress continuity at the layer interfaces and the zero stress condition at the top and bottom of the beam for transverse shear. Numerical examples are presented to illustrate the accuracy of the present model.

  13. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2006-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 2 tabs

  14. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2005-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 3 tabs

  15. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2007-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 2 tabs

  16. Core physics calculation and analysis for SNRE

    International Nuclear Information System (INIS)

    Xie Jiachun; Zhao Shouzhi; Jia Baoshan

    2010-01-01

    Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)

  17. Tank 241-BY-105 rotary core sampling and analysis plan

    International Nuclear Information System (INIS)

    Sasaki, L.M.

    1995-01-01

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two rotary-mode core samples from tank 241-BY-105 (BY-105)

  18. Gas Hydrate-Sediment Morphologies Revealed by Pressure Core Analysis

    Science.gov (United States)

    Holland, M.; Schultheiss, P.; Roberts, J.; Druce, M.

    2006-12-01

    Analysis of HYACINTH pressure cores collected on IODP Expedition 311 and NGHP Expedition 1 showed gas hydrate layers, lenses, and veins contained in fine-grained sediments as well as gas hydrate contained in coarse-grained layers. Pressure cores were recovered from sediments on the Cascadia Margin off the North American West Coast and in the Krishna-Godavari Basin in the Western Bay of Bengal in water depths of 800- 1400 meters. Recovered cores were transferred to laboratory chambers without loss of pressure and nondestructive measurements were made at in situ pressures and controlled temperatures. Gamma density, P-wave velocity, and X-ray images showed evidence of grain-displacing and pore-filling gas hydrate in the cores. Data highlights include X-ray images of fine-grained sediment cores showing wispy subvertical veins of gas hydrate and P-wave velocity excursions corresponding to grain-displacing layers and pore-filling layers of gas hydrate. Most cores were subjected to controlled depressurization experiments, where expelled gas was collected, analyzed for composition, and used to calculate gas hydrate saturation within the core. Selected cores were stored under pressure for postcruise analysis and subsampling.

  19. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  20. Core management and performance analysis for PWR

    International Nuclear Information System (INIS)

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

    1981-01-01

    The KINS (KAERI Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactor fuel management, has been developed and benchmarked against the cycles 1 and 2 of the Kori-1 reactor. The critical boron concentration and three-dimensional power distribution at BOL, HZP condition have been calculated and compared with the operating data. A three-dimensional depletion calculation at HFP condition has been performed for cycle 1 with an interval of 1000 MWD/MTU and compared with the operating data. Similar calculation was also performed for cycle 2 and then compared with the design data of the reactor vendor. At the same time, a prediction of in-core detectors reaction rate was made so as to be compared with the operating data. As the result of comparisons, our calculation as well as the justification of the correlations is shown to be in excellent agreement with the operating data within an allowable limit

  1. Rheumatology training experience across Europe: analysis of core competences.

    Science.gov (United States)

    Sivera, Francisca; Ramiro, Sofia; Cikes, Nada; Cutolo, Maurizio; Dougados, Maxime; Gossec, Laure; Kvien, Tore K; Lundberg, Ingrid E; Mandl, Peter; Moorthy, Arumugam; Panchal, Sonia; da Silva, José A P; Bijlsma, Johannes W

    2016-09-23

    The aim of this project was to analyze and compare the educational experience in rheumatology specialty training programs across European countries, with a focus on self-reported ability. An electronic survey was designed to assess the training experience in terms of self-reported ability, existence of formal education, number of patients managed and assessments performed during rheumatology training in 21 core competences including managing specific diseases, generic competences and procedures. The target population consisted of rheumatology trainees and recently certified rheumatologists across Europe. The relationship between the country of training and the self-reported ability or training methods for each competence was analyzed through linear or logistic regression, as appropriate. In total 1079 questionnaires from 41 countries were gathered. Self-reported ability was high for most competences, range 7.5-9.4 (0-10 scale) for clinical competences, 5.8-9.0 for technical procedures and 7.8-8.9 for generic competences. Competences with lower self-reported ability included managing patients with vasculitis, identifying crystals and performing an ultrasound. Between 53 and 91 % of the trainees received formal education and between 7 and 61 % of the trainees reported limited practical experience (managing ≤10 patients) in each competence. Evaluation of each competence was reported by 29-60 % of the respondents. In adjusted multivariable analysis, the country of training was associated with significant differences in self-reported ability for all individual competences. Even though self-reported ability is generally high, there are significant differences amongst European countries, including differences in the learning structure and assessment of competences. This suggests that educational outcomes may also differ. Efforts to promote European harmonization in rheumatology training should be encouraged and supported.

  2. Size analysis of single-core magnetic nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Frank, E-mail: f.ludwig@tu-bs.de [Institut für Elektrische Messtechnik und Grundlagen der Elektrotechnik, TU Braunschweig, Braunschweig (Germany); Balceris, Christoph; Viereck, Thilo [Institut für Elektrische Messtechnik und Grundlagen der Elektrotechnik, TU Braunschweig, Braunschweig (Germany); Posth, Oliver; Steinhoff, Uwe [Physikalisch-Technische Bundesanstalt, Berlin (Germany); Gavilan, Helena; Costo, Rocio [Instituto de Ciencia de Materiales de Madrid, ICMM/CSIC, Madrid (Spain); Zeng, Lunjie; Olsson, Eva [Department of Applied Physics, Chalmers University of Technology, Göteborg (Sweden); Jonasson, Christian; Johansson, Christer [ACREO Swedish ICT AB, Göteborg (Sweden)

    2017-04-01

    Single-core iron-oxide nanoparticles with nominal core diameters of 14 nm and 19 nm were analyzed with a variety of non-magnetic and magnetic analysis techniques, including transmission electron microscopy (TEM), dynamic light scattering (DLS), static magnetization vs. magnetic field (M-H) measurements, ac susceptibility (ACS) and magnetorelaxometry (MRX). From the experimental data, distributions of core and hydrodynamic sizes are derived. Except for TEM where a number-weighted distribution is directly obtained, models have to be applied in order to determine size distributions from the measurand. It was found that the mean core diameters determined from TEM, M-H, ACS and MRX measurements agree well although they are based on different models (Langevin function, Brownian and Néel relaxation times). Especially for the sample with large cores, particle interaction effects come into play, causing agglomerates which were detected in DLS, ACS and MRX measurements. We observed that the number and size of agglomerates can be minimized by sufficiently strong diluting the suspension. - Highlights: • Investigation of size parameters of single-core magnetic nanoparticles with nominal core diameters of 14 nm and 19 nm utilizing different magnetic and non-magnetic methods • Hydrodynamic size determined from ac susceptibility measurements is consistent with the DLS findings • Core size agrees determined from static magnetization curves, MRX and ACS data agrees with results from TEM although the estimation is based on different models (Langevin function, Brownian and Néel relaxation times).

  3. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1982-07-01

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  4. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J. [Delft University of Technology, NL-2629 JB Delft (Netherlands)

    1998-07-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

  5. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    International Nuclear Information System (INIS)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J.

    1998-01-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

  6. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  7. A core ontology for business process analysis

    NARCIS (Netherlands)

    Pedrinaci, C.; Domingue, J.; Alves De Medeiros, A.K.; Bechhofer, S.; Hauswirth, M.; Hoffmann, J.; Koubarakis, M.

    2008-01-01

    Business Process Management (BPM) aims at supporting the whole life-cycle necessary to deploy and maintain business processes in organisations. An important step of the BPM life-cycle is the analysis of the processes deployed in companies. However, the degree of automation currently achieved cannot

  8. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  9. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    Shatoff, H.; Charman, C.M.

    1983-01-01

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

  10. TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

    Energy Technology Data Exchange (ETDEWEB)

    Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

    2011-07-01

    This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

  11. Sequence comparison and phylogenetic analysis of core gene of ...

    African Journals Online (AJOL)

    Phylogenetic analysis suggests that our sequences are clustered with sequences reported from Japan. This is the first phylogenetic analysis of HCV core gene from Pakistani population. Our sequences and sequences from Japan are grouped into same cluster in the phylogenetic tree. Sequence comparison and ...

  12. Calculation and analysis of generator limiting regimes with respect to stator end core heating

    Directory of Open Access Journals (Sweden)

    Kostić Miloje

    2015-01-01

    Full Text Available A new simplified procedure for defining the limiting operating regimes on the generator capability curve, with respect to stator end core heating, is proposed and described in this paper. First of all, a simplified analysis of axial flux leakage that penetrates into the end plates of the stator is carried out and the corresponding power losses are calculated. Then the analysis of measured point temperature increases over the stator end core, and a qualitative and quantitative overview of the effects, are presented. A simplified procedure for defining the limiting regime with regard to the heating stator end core, which is illustrated for the case of an operating diagram for a given generator of apparent power of 727 MVA (B2 is also described. The given limiting line constructed using this method is similar to the appropriate line constructed on the basis of complex and lengthy factory and on-site tests performed by the manufacturer and the user. According to the results and the check, the proposed method has been proved and the application of the simplified procedure can be recommended for use along with other procedures, at least when it comes to similar synchronous generators in Serbia's Electric Power Industry.

  13. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Science.gov (United States)

    Marks, Janis; Vitolina, Sandra

    2017-12-01

    Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  14. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Directory of Open Access Journals (Sweden)

    Marks Janis

    2017-12-01

    Full Text Available Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  15. Prediction of Hydrophobic Cores of Proteins Using Wavelet Analysis.

    Science.gov (United States)

    Hirakawa; Kuhara

    1997-01-01

    Information concerning the secondary structures, flexibility, epitope and hydrophobic regions of amino acid sequences can be extracted by assigning physicochemical indices to each amino acid residue, and information on structure can be derived using the sliding window averaging technique, which is in wide use for smoothing out raw functions. Wavelet analysis has shown great potential and applicability in many fields, such as astronomy, radar, earthquake prediction, and signal or image processing. This approach is efficient for removing noise from various functions. Here we employed wavelet analysis to smooth out a plot assigned to a hydrophobicity index for amino acid sequences. We then used the resulting function to predict hydrophobic cores in globular proteins. We calculated the prediction accuracy for the hydrophobic cores of 88 representative set of proteins. Use of wavelet analysis made feasible the prediction of hydrophobic cores at 6.13% greater accuracy than the sliding window averaging technique.

  16. Standard Procedure for Grid Interaction Analysis

    International Nuclear Information System (INIS)

    Svensson, Bertil; Lindahl, Sture; Karlsson, Daniel; Joensson, Jonas; Heyman, Fredrik

    2015-01-01

    Grid events, simultaneously affecting all safety related auxiliary systems in a nuclear power plant, are critical and must be carefully addressed in the design, upgrading and operational processes. Up to now, the connecting grid has often been treated as either fully available or totally unavailable, and too little attention has been paid to specify the grid performance criteria. This paper deals with standard procedures for grid interaction analysis, to derive tools and criteria to handle grid events challenging the safety systems of the plant. Critical external power system events are investigated and characterised, with respect to severity and rate of occurrence. These critical events are then grouped with respect to impact on the safety systems, when a disturbance propagates into the plant. It is then important to make sure that 1) the impact of the disturbance will never reach any critical system, 2) the impact of the disturbance will be eliminated before it will hurt any critical system, or 3) the critical systems will be proven to be designed in such a way that they can withstand the impact of the disturbance, and the associated control and protection systems can withstand voltage and frequency transients associated with the disturbances. A number of representative disturbance profiles, reflecting connecting grid conditions, are therefore derived, to be used for equipment testing. (authors)

  17. Identifying functions for ex-core neutron noise analysis

    International Nuclear Information System (INIS)

    Avila, J.M.; Oliveira, J.C.

    1987-01-01

    A method of performing the phase analysis of signals arising from neutron detectors placed in the periphery of a pressurized water reactor is proposed. It consists in the definition of several identifying functions, based on the phases of cross power spectral densities corresponding to four ex-core neutron detectors. Each of these functions enhances the appearance of different sources of noise. The method, applied to the ex-core neutron fluctuation analysis of a French PWR, proved to be very useful as it allows quick recognition of various patterns in the power spectral densities. (orig.) [de

  18. Analysis of Irradiation Holes of In-Core Region

    Energy Technology Data Exchange (ETDEWEB)

    In, Won-ho; Lee, Yong-sub; Kim, Tae-hwan; Lim, Kyoung-hwan; Ahn, Hyung-jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science.

  19. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  20. A trend analysis methodology for enhanced validation of 3-D LWR core simulations

    International Nuclear Information System (INIS)

    Wieselquist, William; Ferroukhi, Hakim; Bernatowicz, Kinga

    2011-01-01

    This paper presents an approach that is being developed and implemented at PSI to enhance the Verification and Validation (V and V) procedure of 3-D static core simulations for the Swiss LWR reactors. The principle is to study in greater details the deviations between calculations and measurements and to assess on that basis if distinct trends of the accuracy can be observed. The presence of such trends could then be a useful indicator of eventual limitations/weaknesses in the applied lattice/core analysis methodology and could thereby serve as guidance for method/model enhancements. Such a trend analysis is illustrated here for a Swiss PWR core model using as basis, the state-of-the-art industrial CASMO/SIMULATE codes. The accuracy of the core-follow models to reproduce the periodic in-core neutron flux measurements is studied for a total of 21 operating cycles. The error is analyzed with respect to different physics parameters with a ranking of the individual assemblies/nodes contribution to the total RMS error and trends are analyzed by performing partial correlation analysis. The highest errors appear at the core axial peripheries (top/bottom nodes) where a mean C/E-1 error of 10% is observed for the top nodes and -5% for the bottom nodes and the maximum C/E-1 error reaches almost 20%. Partial correlation analysis shows significant correlation of error to distance from core mid-plane and only less significant correlations to other variables. Overall, it appears that the primary areas that could benefit from further method/modeling improvements are: axial reflectors, MOX treatment and control rod cusping. (author)

  1. A retrospective analysis of ultrasound-guided large core needle ...

    African Journals Online (AJOL)

    A retrospective analysis of ultrasound-guided large core needle biopsies of breast lesions at a regional public hospital in Durban, KwaZulu-Natal, South Africa. ... Objective: To assess the influence of technical variables on the diagnostic yield of breast specimens obtained by using US-LCNB, and the sensitivity of detecting ...

  2. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  3. STYCA, a computer program in the dynamic structural analysis of a PWR core

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da; Breyne Salvagni, R. de

    1992-01-01

    A procedure for the dynamic structural analysis of a PWR core is presented, impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. A time-history response analysis is necessary. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. An algorithm of solution and also results obtained with the STYCA computer program, developed on the basis of what was proposed here, are presented. (author)

  4. Dynamic structural analysis for assemblies of fuel elements in the core of a PWR

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da.

    1991-01-01

    It is presented a procedure for the dynamic structural analysis of a PWR core. Impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. It is necessary a time-history response analysis. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. It is presented an algorithm of solution and also results obtained with the STYCA computer program, developed in the basis of what was proposed here. (author)

  5. In-core LOCA (PTR) analysis with poisoned moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

    2005-01-01

    CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

  6. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  7. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  8. Refurbishment, core conversion and safety analysis of Apsara reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

    1998-07-01

    Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

  9. Summative Mass Analysis of Algal Biomass - Integration of Analytical Procedures: Laboratory Analytical Procedure (LAP)

    Energy Technology Data Exchange (ETDEWEB)

    Laurens, Lieve M. L.

    2016-01-13

    This procedure guides the integration of laboratory analytical procedures to measure algal biomass constituents in an unambiguous manner and ultimately achieve mass balance closure for algal biomass samples. Many of these methods build on years of research in algal biomass analysis.

  10. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Watanabe, Norio

    1982-06-01

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10 - 3 /demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  11. Scoping Analysis on Core Disruptive Accident in PGSFR (2015 Results)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In general, the severe accident is classified by three phases. The first phase is the initiation (pre-disassembly) phase that occurs the gradual core meltdown from accident initiation to the point of neutronic shutdown with an intact geometry. The second phase is the transition phase that happens the fuel transition from a solid to a liquid phase. Fuel and cladding can melt to form a molten pool and core can boil, then criticality conditions can recur. The third phase is the disassembly phase. In other words, this phase is Core Disruptive Accident (CDA). Power excursion is followed until the core is disassembled in this phase. In the early considerations of Liquid Metal Fast Breeder Reactor (LMFBR) energetics, the term Hypothetical Core Disruptive Accidents (HCDAs) was in common use. This was not only to connote the extremely low probability of initiation of such accidents, but also the tentative nature of our understanding of their behavior and resulting consequences. A numerical analysis is conducted to estimate the energy release, pressure behavior and core expansion behavior induced by CDA of PGSFR using CDA-ER and CDA-CEME codes. Conservatively, the calculated results of energy release and pressure behavior induced by CDA without Doppler effect in PGSFR when whole cores were melted (100 $/s) were 7.844 GJ and 4.845 GPa, respectively. With Doppler effect, the analyzed maximum energy release and pressure were 6.696 GJ and 3.449 GPa, respectively. The calculated results of the core expansion behavior during 0.015 seconds after the explosion without Doppler effect in PGSFR when whole cores were melted (100 $/s) were as follows: The total energy is calculated to be 1.87 GJ. At 0.01 s, the kinetic energy of the sodium is 1.85 GJ, while the expansion work and internal energy of the bubble are 19.7 MJ and 0.98 J, respectively. With Doppler effect, the total energy is calculated to be 1.33 GJ. At 0.01 s, the kinetic energy of the sodium is 1.31 GJ, while the expansion

  12. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  13. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  14. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  15. Monte carlo depletion analysis of SMART core by MCNAP code

    International Nuclear Information System (INIS)

    Jung, Jong Sung; Sim, Hyung Jin; Kim, Chang Hyo; Lee, Jung Chan; Ji, Sung Kyun

    2001-01-01

    Depletion an analysis of SMART, a small-sized advanced integral PWR under development by KAERI, is conducted using the Monte Carlo (MC) depletion analysis program, MCNAP. The results are compared with those of the CASMO-3/ MASTER nuclear analysis. The difference between MASTER and MCNAP on k eff prediction is observed about 600pcm at BOC, and becomes smaller as the core burnup increases. The maximum difference bet ween two predict ions on fuel assembly (FA) normalized power distribution is about 6.6% radially , and 14.5% axially but the differences are observed to lie within standard deviation of MC estimations

  16. Solid-phase extraction procedures in systematic toxicological analysis

    NARCIS (Netherlands)

    Franke, J.P.; de Zeeuw, R.A

    1998-01-01

    In systematic toxicological analysis (STA) the substance(s) present is (are) not known at the start of the analysis. in such an undirected search the extraction procedure cannot be directed to a given substance but must be a general procedure where a compromise must be reached in that the substances

  17. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  18. 76 FR 5651 - Practice and Procedure; Amendment of CORES Registration System

    Science.gov (United States)

    2011-02-01

    ... agencies to establish procedures to identify the causes of overpayments, delinquencies, and defaults, and... review the progress on their filings, fees that are due, the history of files the filer has submitted, as...

  19. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  20. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  1. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo

    1999-01-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  2. Evaluation of a thermal SCWR core with sub-channel analysis

    International Nuclear Information System (INIS)

    Liu Xiaojing; Cheng Xu

    2008-01-01

    A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

  3. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-15

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

  4. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    International Nuclear Information System (INIS)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-01

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty

  5. Analysis procedure for americium in environmental samples

    International Nuclear Information System (INIS)

    Holloway, R.W.; Hayes, D.W.

    1982-01-01

    Several methods for the analysis of 241 Am in environmental samples were evaluated and a preferred method was selected. This method was modified and used to determine the 241 Am content in sediments, biota, and water. The advantages and limitations of the method are discussed. The method is also suitable for 244 Cm analysis

  6. Severe core damage experiments and analysis for CANDU applications

    International Nuclear Information System (INIS)

    Mathew, P.M.; White, A.J.; Snell, V.G.; Bonechi, M.

    2003-01-01

    AECL uses the MAAP CANDU code to calculate the progression of a severe core damage accident in a CANDU reactor to support Level 2 Probabilistic Safety Assessment and Severe Accident Management activities. Experimental data are required to ensure that the core damage models used in MAAP CANDU code are adequate. In SMiRT 16, details of single channel experiments were presented to elucidate the mechanisms of core debris formation. This paper presents the progress made in severe core damage experiments since then using single channels in an inert atmosphere and results of the model development work to support the experiments. The core disassembly experiments are conducted with one-fifth scale channels made of Zr-2.5wt%Nb containing twelve simulated fuel bundles in an inert atmosphere. The reference fuel channel geometry consists of a pressure tube/calandria tube composite, with the pressure tube ballooned into circumferential contact with the calandria tube. Experimental results from single channel tests showed the development of time-dependent sag when the reference channel temperature exceeded 850 degC. The test results also showed significant strain localization in the gap at the bundle junctions along the bottom side of the channel, thus suggesting creep to be the main deformation mechanism for debris formation. An ABAQUS finite element model using two-dimensional beam elements with circular cross-section was developed to explain the experimental findings. A comparison of the calculated central sag (at mid-span), the axial displacement at the free end of the channel and the post-test sag profile showed good agreement with the experiments, when strain localization was included in the model, suggesting such a simple modelling approach would be adequate to explain the test findings. The results of the tests are important not only in the context of the validation of the analytical tools and models adopted by AECL for the severe accident analysis of CANDU reactors but

  7. High enrichment to low enrichment core's conversion. Accidents analysis

    International Nuclear Information System (INIS)

    Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

    1990-01-01

    This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

  8. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  9. Building America Performance Analysis Procedures: Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    2004-06-01

    To measure progress toward multi-year research goals, cost and performance trade-offs are evaluated through a series of controlled field and laboratory experiments supported by energy analysis techniques using test data to calibrate simulation models.

  10. Core disruptive accident and recriticality analysis with FX2-POOL

    International Nuclear Information System (INIS)

    Abramson, P.B.

    1976-01-01

    The current state of development of FX2-POOL, a two-dimensional hydrodynamic, thermodynamic and neutronic scoping model for Hypothetical Core Disruptive Accident analysis is described. Checkout comparisons to VENUS for prompt burst conditions were good. Use of FX2-POOL to examine the importance of fuel to steel heat transfer during a prompt burst indicates that heat transfer plays no important role on that time scale. Scoping studies of material thermohydrodynamics for about 20 to 30 milliseconds following the prompt burst indicate that heat transfer is important on the time scale necessary for the CDA bubble to grow to the size of the original core. Preliminary results are presented for energetics of boiling fuel steel pools which are forced recritical by local surface pressurization

  11. Analysis of fetal dose in CT procedures

    International Nuclear Information System (INIS)

    Ortiz Torres, A.; Plazas, M. C.

    2006-01-01

    It is the miracle of the life, that sublime formation, the given more beautiful gift for heaven's sake to our to exist, and it is consequently our responsibility to look after their protection and care. Today in day the quantity of radiation absorbed by the fetus in the treatments for radiodiagnostic, mainly in the procedures of on-line axial tomography, the fetus absorbs a considerable dose of radiation and the questions generated regarding if these doses, bear to a risk of malformations or if it is necessary the interruption of the pregnancy is very frequent. In most of the cases, the treatment with ionizing radiations that it is beneficial for the mother, is only indirectly it for the fetus that is exposed to a risk. The possibility that a fetus or a small boy contract cancer caused by the radiation it can be three times superior to that of the population in general, of there the importance of analyzing the goods of the prenatal irradiation and the main agents to consider for the estimate of the magnitude of the risk of the exhibitions in uterus. In the different circumstances in that these can happen in treatments of on-line axial tomography computerized. (Author)

  12. Shakedown analysis by finite element incremental procedures

    International Nuclear Information System (INIS)

    Borkowski, A.; Kleiber, M.

    1979-01-01

    It is a common occurence in many practical problems that external loads are variable and the exact time-dependent history of loading is unknown. Instead of it load is characterized by a given loading domain: a convex polyhedron in the n-dimensional space of load parameters. The problem is then to check whether a structure shakes down, i.e. responds elastically after a few elasto-plastic cycles, or not to a variable loading as defined above. Such check can be performed by an incremental procedure. One should reproduce incrementally a simple cyclic process which consists of proportional load paths that connect the origin of the load space with the corners of the loading domain. It was proved that if a structure shakes down to such loading history then it is able to adopt itself to an arbitrary load path contained in the loading domain. The main advantage of such approach is the possibility to use existing incremental finite-element computer codes. (orig.)

  13. Using a Simultaneous Prompting Procedure to Embed Core Content When Teaching a Potential Employment Skill

    Science.gov (United States)

    Collins, Belva C.; Terrell, Misty; Test, David W.

    2017-01-01

    This investigation used a multiple-probe-across-participants design to examine the effects of using a simultaneous prompting procedure to teach four secondary students with mild intellectual disabilities the employment task of caring for plants in a greenhouse. The instructor also embedded photosynthesis science content as nontargeted information…

  14. Video content analysis of surgical procedures.

    Science.gov (United States)

    Loukas, Constantinos

    2018-02-01

    In addition to its therapeutic benefits, minimally invasive surgery offers the potential for video recording of the operation. The videos may be archived and used later for reasons such as cognitive training, skills assessment, and workflow analysis. Methods from the major field of video content analysis and representation are increasingly applied in the surgical domain. In this paper, we review recent developments and analyze future directions in the field of content-based video analysis of surgical operations. The review was obtained from PubMed and Google Scholar search on combinations of the following keywords: 'surgery', 'video', 'phase', 'task', 'skills', 'event', 'shot', 'analysis', 'retrieval', 'detection', 'classification', and 'recognition'. The collected articles were categorized and reviewed based on the technical goal sought, type of surgery performed, and structure of the operation. A total of 81 articles were included. The publication activity is constantly increasing; more than 50% of these articles were published in the last 3 years. Significant research has been performed for video task detection and retrieval in eye surgery. In endoscopic surgery, the research activity is more diverse: gesture/task classification, skills assessment, tool type recognition, shot/event detection and retrieval. Recent works employ deep neural networks for phase and tool recognition as well as shot detection. Content-based video analysis of surgical operations is a rapidly expanding field. Several future prospects for research exist including, inter alia, shot boundary detection, keyframe extraction, video summarization, pattern discovery, and video annotation. The development of publicly available benchmark datasets to evaluate and compare task-specific algorithms is essential.

  15. Analysis and optimization of blood-testing procedures.

    NARCIS (Netherlands)

    Bar-Lev, S.K.; Boxma, O.J.; Perry, D.; Vastazos, L.P.

    2017-01-01

    This paper is devoted to the performance analysis and optimization of blood testing procedures. We present a queueing model of two queues in series, representing the two stages of a blood-testing procedure. Service (testing) in stage 1 is performed in batches, whereas it is done individually in

  16. System analysis procedures for conducting PSA of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Yoon Hwan; Jeong, Won Dae; Kim, Tae Un; Kim, Kil You; Han, Sang Hoon; Chang, Seung Chul; Sung, Tae Yong; Yang, Jun Eon; Kang, Dae Il; Park, Jin Hee; Hwang, Mi Jeong; Jin, Young Ho.

    1997-03-01

    This document, the Probabilistic Safety Assessment(PSA) procedures guide for system analysis, is intended to provide the guidelines to analyze the target of system consistently and technically in the performance of PSA for nuclear power plants(NPPs). The guide has been prepared in accordance with the procedures and techniques for fault tree analysis(FTA) used in system analysis. Normally the main objective of system analysis is to assess the reliability of system modeled by Event Tree Analysis(ETA). A variety of analytical techniques can be used for the system analysis, however, FTA method is used in this procedures guide. FTA is the method used for representing the failure logic of plant systems deductively using AND, OR or NOT gates. The fault tree should reflect all possible failure modes that may contribute to the system unavailability. This should include contributions due to the mechanical failures of the components, Common Cause Failures (CCFs), human errors and outages for testing and maintenance. After the construction of fault tree is completed, system unavailability is calculated with the CUT module of KIRAP, and the qualitative and quantitative analysis is performed through the process as above stated. As above mentioned, the procedures for system analysis is based on PSA procedures and methods which has been applied to the safety assessments of constructing NPPs in the country. Accordingly, the method of FTA stated in this procedures guide will be applicable to PSA for the NPPs to be constructed in the future. (author). 6 tabs., 11 figs., 7 refs

  17. System analysis procedures for conducting PSA of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hwan; Jeong, Won Dae; Kim, Tae Un; Kim, Kil You; Han, Sang Hoon; Chang, Seung Chul; Sung, Tae Yong; Yang, Jun Eon; Kang, Dae Il; Park, Jin Hee; Hwang, Mi Jeong; Jin, Young Ho

    1997-03-01

    This document, the Probabilistic Safety Assessment(PSA) procedures guide for system analysis, is intended to provide the guidelines to analyze the target of system consistently and technically in the performance of PSA for nuclear power plants(NPPs). The guide has been prepared in accordance with the procedures and techniques for fault tree analysis(FTA) used in system analysis. Normally the main objective of system analysis is to assess the reliability of system modeled by Event Tree Analysis(ETA). A variety of analytical techniques can be used for the system analysis, however, FTA method is used in this procedures guide. FTA is the method used for representing the failure logic of plant systems deductively using AND, OR or NOT gates. The fault tree should reflect all possible failure modes that may contribute to the system unavailability. This should include contributions due to the mechanical failures of the components, Common Cause Failures (CCFs), human errors and outages for testing and maintenance. After the construction of fault tree is completed, system unavailability is calculated with the CUT module of KIRAP, and the qualitative and quantitative analysis is performed through the process as above stated. As above mentioned, the procedures for system analysis is based on PSA procedures and methods which has been applied to the safety assessments of constructing NPPs in the country. Accordingly, the method of FTA stated in this procedures guide will be applicable to PSA for the NPPs to be constructed in the future. (author). 6 tabs., 11 figs., 7 refs.

  18. Neutronic analysis of LMFBRs during severe core disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1979-01-01

    A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction

  19. Quantification procedures in micro X-ray fluorescence analysis

    International Nuclear Information System (INIS)

    Kanngiesser, Birgit

    2003-01-01

    For the quantification in micro X-ray fluorescence analysis standardfree quantification procedures have become especially important. An introduction to the basic concepts of these quantification procedures is given, followed by a short survey of the procedures which are available now and what kind of experimental situations and analytical problems are addressed. The last point is extended by the description of an own development for the fundamental parameter method, which renders the inclusion of nonparallel beam geometries possible. Finally, open problems for the quantification procedures are discussed

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

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

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

  1. Training the Neglected Core of Army Leadership- Troop-Leading Procedures

    Science.gov (United States)

    2007-06-01

    tactical operations center ( TOC ) to reconnoiter, using the MDMP to complete their plans, etc. TLP are considered procedures, and the Army’s previous...and establishing an effective and responsive Prevention of Sexual Haras- sment (POSH)/Equal Opportunity (EO) program. Since dedicated citizen...8. One unit posted signs all over its TOC that read, “Who else needs to know!” 03A - COE/Full-Spectrum Operations/Why We Fight 03C - Perform Cultural

  2. Operating procedures: Fusion Experiments Analysis Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lerche, R.A.; Carey, R.W.

    1984-03-20

    The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility.

  3. Building America Performance Analysis Procedures: Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hendron, R.; Anderson, R.; Judkoff, R.; Christensen, C.; Eastment, M.; Norton, P.; Reeves, P.; Hancock, E.

    2004-06-01

    To measure progress toward multi-year Building America research goals, cost and performance trade-offs are evaluated through a series of controlled field and laboratory experiments supported by energy analysis techniques that use test data to''calibrate'' energy simulation models. This report summarizes the guidelines for reporting such analytical results using the Building America Research Benchmark (Version 3.1) in studies that also include consideration of current Regional and Builder Standard Practice. Version 3.1 of the Benchmark is generally consistent with the 1999 Home Energy Rating System (HERS) Reference Home, with additions that allow evaluation of all home energy uses.

  4. Operating procedures: Fusion Experiments Analysis Facility

    International Nuclear Information System (INIS)

    Lerche, R.A.; Carey, R.W.

    1984-01-01

    The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility

  5. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  6. Development of local TDC model in core thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Kwon, H.S.; Park, J.R.; Hwang, D.H.; Lee, S.K.

    2004-01-01

    The local TDC model consisting of natural mixing and forced mixing part was developed to obtain more realistic local fluid properties in the core subchannel analysis. To evaluate the performance of local TDC model, the CHF prediction capability was tested with the various CHF correlations and local fluid properties at CHF location which are based on the local TDC model. The results show that the standard deviation of measured to predicted CHF ratio (M/P) based on local TDC model can be reduced by about 7% compared to those based on global TDC model when the CHF correlation has no term to account for distance from the spacer grid. (author)

  7. Case for integral core-disruptive accident analysis

    International Nuclear Information System (INIS)

    Luck, L.B.; Bell, C.R.

    1985-01-01

    Integral analysis is an approach used at the Los Alamos National Laboratory to cope with the broad multiplicity of accident paths and complex phenomena that characterize the transition phase of core-disruptive accident progression in a liquid-metal-cooled fast breeder reactor. The approach is based on the combination of a reference calculation, which is intended to represent a band of similar accident paths, and associated system- and separate-effect studies, which are designed to determine the effect of uncertainties. Results are interpreted in the context of a probabilistic framework. The approach was applied successfully in two studies; illustrations from the Clinch River Breeder Reactor licensing assessment are included

  8. 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.

  9. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    Nicolas, Anne

    1989-01-01

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated [fr

  10. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  11. Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2008-01-01

    The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

  12. Analysis of core damage frequency, Surry, Unit 1 internal events appendices

    International Nuclear Information System (INIS)

    Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

    1990-04-01

    This document contains the appendices for the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed is an extensive reanalysis of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency

  13. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  14. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  15. Analysis the Response Function of the HTR Ex-core Neutron Detectors in Different Core Status

    International Nuclear Information System (INIS)

    Fan Kai; Li Fu; Zhou Xuhua

    2014-01-01

    Modular high temperature gas cooled reactor HTR-PM demonstration plant, designed by INET, Tsinghua University, is being built in Shidao Bay, Shandong province, China. HTR-PM adopts pebble bed concept. The harmonic synthesis method has been developed to reconstruct the power distributions on HTR-PM. The method based on the assumption that the neutron detector readings are mainly determined by the status of the core through the power distribution, and the response functions changed little when the status of the core changed. To verify the assumption, the influence factors to the ex-core neutron detectors are calculated in this paper, including the control rod position and the temperature of the core. The results shows that when the status of the core changed, the power distribution changed more remarkable than the response function, but the detector readings could change about 5% because of the response function changing. (author)

  16. Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

    International Nuclear Information System (INIS)

    Itagaki, M.; Abe, J.I.; Kuribayashi, K.

    1987-01-01

    There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

  17. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational challenge in staging...... a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have developed and refined...

  18. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  19. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  20. Current Human Reliability Analysis Methods Applied to Computerized Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring

    2012-06-01

    Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room (Fink et al., 2009). Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of enhanced ease of use and easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

  1. Application of Network Analysis Method to VHTR core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl

    2012-01-01

    A Very High Temperature Reactor (VHTR) is currently envisioned as a promising future reactor concept because of its high-efficiency and capability of generating hydrogen. Prismatic Modular Reactor (PMR) is one of the main VHTR concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However their shape could be changed by neutron damage during the reactor operation and the shape change can makes the gaps between the blocks inducing bypass flow. Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Therefore, fast, flexible and reliable code is required to predict the flow distribution corresponding to the various bypass gap distribution. Consequently in this study, the flow network analysis method is applied to analyze the core flow of VHTR. The applied method was validated by comparing with SNU VHTR multiblock experiment. As a result, the calculated results show good agreements with experimental data although computational time and cost of the developed code was very small

  2. Experimental programme and analysis, ZENITH II, Core 4

    Energy Technology Data Exchange (ETDEWEB)

    Ingram, G.; Sanders, J. E.; Sherwin, J.

    1974-10-15

    The Phase 3 program of reactor physics experiments on the HTR (or Mk 3 GCR) lattices continued during the first half of 1974 with a study of a series of critical builds in Zenith II aimed at testing predictions of shut-down margins in the local criticality-situations arising during power reactor refueling. The paper describes the experimental program and the subsequent theoretical analysis using methods developed in the United Kingdom for calculating low-enriched uranium HTR fuel systems. The importance of improving the accuracy of predictions of shut-down margins arises from the basic requirement that the core in its most reactive condition and with a specified number of absorbers removed from the array must remain sub-critical with a margin adequate to cover the total uncertainty of +/- 1 Nile (that is, 1 % delta-k). The major uncertainty is that in modelling the complex fuel/absorber configuration, and this is the aspect essentially covered in the Zenith II Core 4 studies.

  3. Improvement of numerical analysis method for FBR core characteristics. 3

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yamamoto, Toshihisa; Kitada, Takanori; Katagi, Yousuke

    1998-03-01

    As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method; A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. Part 2: Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory; Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. Part 3: Improvement of Nodal Transport Calculation for Hexagonal Geometry; A method to evaluate the intra-subassembly power distribution from the nodal averaged neutron flux and surface fluxes at the node boundaries, was developed based on the transport theory. (J.P.N.)

  4. Procedure for the analysis of americium in complex matrices

    International Nuclear Information System (INIS)

    Knab, D.

    1978-02-01

    A radioanalytical procedure for the analysis of americium in complex matrices has been developed. Clean separations of americium can be obtained from up to 100 g of sample ash, regardless of the starting material. The ability to analyze large masses of material provides the increased sensitivity necessary to detect americium in many environmental samples. The procedure adequately decontaminates from rare earth elements and natural radioactive nuclides that interfere with the alpha spectrometric measurements

  5. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  6. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  7. Accident Sequence Evaluation Program: Human reliability analysis procedure

    International Nuclear Information System (INIS)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

  8. FBR core mock-up RAPSODIE I - experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with and without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests. (author)

  9. Powered bone marrow biopsy procedures produce larger core specimens, with less pain, in less time than with standard manual devices

    Directory of Open Access Journals (Sweden)

    Larry J. Miller

    2011-07-01

    Full Text Available Bone marrow sampling remains essential in the evaluation of hematopoietic and many non-hematopoietic disorders. One common limitation to these procedures is the discomfort experienced by patients. To address whether a Powered biopsy system could reduce discomfort while providing equivalent or better results, we performed a randomized trial in adult volunteers. Twenty-six subjects underwent bilateral biopsies with each device. Core samples were obtained in 66.7% of Manual insertions; 100% of Powered insertions (P=0.002. Initial mean biopsy core lengths were 11.1±4.5 mm for the Manual device; 17.0±6.8 mm for the Powered device (P<0.005. Pathology assessment for the Manual device showed a mean length of 6.1±5.6 mm, width of 1.0±0.7 mm, and volume of 11.0±10.8 mm3. Powered device measurements were mean length of 15.3±6.1 mm, width of 2.0±0.3 mm, and volume of 49.1±21.5 mm3 (P<0.001. The mean time to core ejection was 86 seconds for Manual device; 47 seconds for the Powered device (P<0.001. The mean second look overall pain score was 33.3 for the Manual device; 20.9 for the Powered (P=0.039. We conclude that the Powered biopsy device produces superior sized specimens, with less overall pain, in less time.

  10. Automated software analysis of nuclear core discharge data

    International Nuclear Information System (INIS)

    Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

    1993-03-01

    Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

  11. Development and analysis of U-core switched reluctance machine

    DEFF Research Database (Denmark)

    Jæger, Rasmus; Nielsen, Simon Staal; Rasmussen, Peter Omand

    2016-01-01

    Switched reluctance machines (SRMs) have seen a lot of interest due to their rugged and fault tolerant construction as well as their high efficiency over a wide speed range. The technology however suffers from torque ripple, acoustic noise and low torque density. Many concepts to address these di......Switched reluctance machines (SRMs) have seen a lot of interest due to their rugged and fault tolerant construction as well as their high efficiency over a wide speed range. The technology however suffers from torque ripple, acoustic noise and low torque density. Many concepts to address...... and reduced flux reversal, reducing core losses. Due to an increased number of poles, torque density is increased and torque ripple reduced. A prototype is built and through a number of tests, the machine is mapped and all loss components are analysed. As a result of the analysis, an assessment is presented...

  12. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

    1992-10-01

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  13. A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2005-02-01

    In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

  14. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

    1996-12-01

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

  15. Analysis of hypothetical nuclear excursions in the external core retention system

    International Nuclear Information System (INIS)

    Froehlich, R.; Kussmaul, G.; Schmuck, P.

    1976-01-01

    The core catcher system of the SNR 300 is outside the reactor tank. The probability of recriticality phenomena is reduced by its design, but the licensing procedures still call for the analysis of strong recriticality phenomena in the core catcher system outside the reactor tank in order to achieve a better understanding of the possible physical effects and to get to know the safety limits of the system. For their theoretical investigations, the authors used a two-partner model as presented in fig. 1. At the bottom of the core catcher - which consists of depleted UO 2 - there is a fuel cylinder. Another fuel cylinder (with the same axis) is dropped from a height of 250 cm. The two cylindrical masses are immersed in sodium, but a free fall is assumed since the possibility cannot be excluded that the reactor bottom may be empty or only partially filled with sodium. It was found that under these conditions the strongest excursions may be expected in those cases where prompt criticality does not occur until just before the two partners meet. (orig./AK) [de

  16. Procedures for the external event core damage frequency analyses for NUREG-1150

    International Nuclear Information System (INIS)

    Bohn, M.P.; Lambright, J.A.

    1990-11-01

    This report presents methods which can be used to perform the assessment of risk due to external events at nuclear power plants. These methods were used to perform the external events risk assessments for the Surry and Peach Bottom nuclear power plants as part of the NRC-sponsored NUREG-1150 risk assessments. These methods apply to the full range of hazards such as earthquakes, fires, floods, etc. which are collectively known as external events. The methods described in this report have been developed under NRC sponsorship and represent, in many cases, both advancements and simplifications over techniques that have been used in past years. They also include the most up-to-date data bases on equipment seismic fragilities, fire occurrence frequencies and fire damageability thresholds. The methods described here are based on making full utilization of the power plant systems logic models developed in the internal events analyses. By making full use of the internal events models one obtains an external event analysis that is consistent both in nomenclature and in level of detail with the internal events analyses, and in addition, automatically includes all the appropriate random and tests/maintenance unavailabilities as appropriate. 50 refs., 9 figs., 11 tabs

  17. Method and procedure of fatigue analysis for nuclear equipment

    International Nuclear Information System (INIS)

    Wen Jing; Fang Yonggang; Lu Yan; Zhang Yue; Sun Zaozhan; Zou Mingzhong

    2014-01-01

    As an example, the fatigue analysis for the upper head of the pressurizer in one NPP was carried out by using ANSYS, a finite element method analysis software. According to RCC-M code, only two kinds of typical transients of temperature and pressure were considered in the fatigue analysis. Meanwhile, the influence of earthquake was taken into account. The method and procedure of fatigue analysis for nuclear safety equipment were described in detail. This paper provides a reference for fatigue analysis and assessment of nuclear safety grade equipment and pipe. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-15

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

  19. Procedure for conducting probabilistic safety assessment: level 1 full power internal event analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Won Dae; Lee, Y. H.; Hwang, M. J. [and others

    2003-07-01

    This report provides guidance on conducting a Level I PSA for internal events in NPPs, which is based on the method and procedure that was used in the PSA for the design of Korea Standard Nuclear Plants (KSNPs). Level I PSA is to delineate the accident sequences leading to core damage and to estimate their frequencies. It has been directly used for assessing and modifying the system safety and reliability as a key and base part of PSA. Also, Level I PSA provides insights into design weakness and into ways of preventing core damage, which in most cases is the precursor to accidents leading to major accidents. So Level I PSA has been used as the essential technical bases for risk-informed application in NPPs. The report consists six major procedural steps for Level I PSA; familiarization of plant, initiating event analysis, event tree analysis, system fault tree analysis, reliability data analysis, and accident sequence quantification. The report is intended to assist technical persons performing Level I PSA for NPPs. A particular aim is to promote a standardized framework, terminology and form of documentation for PSAs. On the other hand, this report would be useful for the managers or regulatory persons related to risk-informed regulation, and also for conducting PSA for other industries.

  20. 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.

  1. Similarity and uncertainty analysis of the ALLEGRO MOX core

    International Nuclear Information System (INIS)

    Vrban, B.; Hascik, J.; Necas, V.; Slugen, V.

    2015-01-01

    The similarity and uncertainty analysis of the ESNII+ ALLEGRO MOX core has identified specific problems and challenges in the field of neutronic calculations. Similarity assessment identified 9 partly comparable experiments where only one reached ck and E values over 0.9. However the Global Integral Index G remains still low (0.75) and cannot be judge das sufficient. The total uncertainty of calculated k eff induced by XS data is according to our calculation 1.04%. The main contributors to this uncertainty are 239 Pu nubar and 238 U inelastic scattering. The additional margin from uncovered sensitivities was determined to be 0.28%. The identified low number of similar experiments prevents the use of advanced XS adjustment and bias estimation methods. More experimental data are needed and presented results may serve as a basic step in development of necessary critical assemblies. Although exact data are not presented in the paper, faster 44 energy group calculation gives almost the same results in similarity analysis in comparison to more complex 238 group calculation. Finally, it was demonstrated that TSUNAMI-IP utility can play a significant role in the future fast reactor development in Slovakia and in the Visegrad region. Clearly a further Research and Development and strong effort should be carried out in order to receive more complex methodology consisting of more plausible covariance data and related quantities. (authors)

  2. Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

    International Nuclear Information System (INIS)

    Ohtaka, Masahiko; Ohshima, Hiroyuki

    1998-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

  3. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    International Nuclear Information System (INIS)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L.

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs

  4. ORNL: PWR-BDHT analysis procedure, a preliminary overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The computer programs currently used in the analysis of the ORNL-PWR Blowdown Heat Transfer Separate-Effects Program are overviewed. The current linkages and relationships among the programs are given along with general comments about the future directions of some of these programs. The overview is strictly from the computer science point of view with only minimal information concerning the engineering aspects of the analysis procedure

  5. Establishment of analysis procedure for control rod reactivity worth

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

    2001-03-01

    As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on.

  6. Establishment of analysis procedure for control rod reactivity worth

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

    2001-03-01

    As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on

  7. Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

    International Nuclear Information System (INIS)

    Shin, Hyeong-Ki

    1999-01-01

    The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

  8. Cost analysis of robotic versus laparoscopic general surgery procedures.

    Science.gov (United States)

    Higgins, Rana M; Frelich, Matthew J; Bosler, Matthew E; Gould, Jon C

    2017-01-01

    Robotic surgical systems have been used at a rapidly increasing rate in general surgery. Many of these procedures have been performed laparoscopically for years. In a surgical encounter, a significant portion of the total costs is associated with consumable supplies. Our hospital system has invested in a software program that can track the costs of consumable surgical supplies. We sought to determine the differences in cost of consumables with elective laparoscopic and robotic procedures for our health care organization. De-identified procedural cost and equipment utilization data were collected from the Surgical Profitability Compass Procedure Cost Manager System (The Advisory Board Company, Washington, DC) for our health care system for laparoscopic and robotic cholecystectomy, fundoplication, and inguinal hernia between the years 2013 and 2015. Outcomes were length of stay, case duration, and supply cost. Statistical analysis was performed using a t-test for continuous variables, and statistical significance was defined as p robotic procedures. Length of stay did not differ for fundoplication or cholecystectomy. Length of stay was greater for robotic inguinal hernia repair. Case duration was similar for cholecystectomy (84.3 robotic and 75.5 min laparoscopic, p = 0.08), but significantly longer for robotic fundoplication (197.2 robotic and 162.1 min laparoscopic, p = 0.01) and inguinal hernia repair (124.0 robotic and 84.4 min laparoscopic, p = ≪0.01). We found a significantly increased cost of general surgery procedures for our health care system when cases commonly performed laparoscopically are instead performed robotically. Our analysis is limited by the fact that we only included costs associated with consumable surgical supplies. The initial acquisition cost (over $1 million for robotic surgical system), depreciation, and service contract for the robotic and laparoscopic systems were not included in this analysis.

  9. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  10. Methodologies for uncertainty analysis in the level 2 PSA and their implementation procedures

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Yang, Joon Eun; Kim, Dong Ha

    2002-04-01

    Main purpose of this report to present standardized methodologies for uncertainty analysis in the Level 2 Probabilistic Safety Assessment (PSA) and their implementation procedures, based on results obtained through a critical review of the existing methodologies for the analysis of uncertainties employed in the Level 2 PSA, especially Accident Progression Event Tree (APET). Uncertainties employed in the Level 2 PSA, quantitative expressions of overall knowledge of analysts' and experts' participating in the probabilistic quantification process of phenomenological accident progressions ranging from core melt to containment failure, their numerical values are directly related to the degree of confidence that the analyst has that a given phenomenological event or accident process will or will not occur, or analyst's subjective probabilities of occurrence. These results that are obtained from Level 2 PSA uncertainty analysis, become an essential contributor to the plant risk, in addition to the Level 1 PSA and Level 3 PSA uncertainties. Uncertainty analysis methodologies and their implementation procedures presented in this report was prepared based on the following criteria: 'uncertainty quantification process must be logical, scrutable, complete, consistent and in an appropriate level of detail, as mandated by the Level 2 PSA objectives'. For the aforementioned purpose, this report deals mainly with (1) summary of general or Level 2 PSA specific uncertainty analysis methodologies, (2) selection of phenomenological branch events for uncertainty analysis in the APET, methodology for quantification of APET uncertainty inputs and its implementation procedure, (3) statistical propagation of uncertainty inputs through APET and its implementation procedure, and (4) formal procedure for quantification of APET uncertainties and source term categories (STCs) through the Level 2 PSA quantification codes

  11. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  12. Flood risk analysis procedure for nuclear power plants

    International Nuclear Information System (INIS)

    Wagner, D.P.

    1982-01-01

    This paper describes a methodology and procedure for determining the impact of floods on nuclear power plant risk. The procedures are based on techniques of fault tree and event tree analysis and use the logic of these techniques to determine the effects of a flood on system failure probability and accident sequence occurrence frequency. The methodology can be applied independently or as an add-on analysis for an existing risk assessment. Each stage of the analysis yields useful results such as the critical flood level, failure flood level, and the flood's contribution to accident sequence occurrence frequency. The results of applications show the effects of floods on the risk from nuclear power plants analyzed in the Reactor Safety Study

  13. Sample preparation procedure for PIXE elemental analysis on soft tissues

    International Nuclear Information System (INIS)

    Kubica, B.; Kwiatek, W.M.; Dutkiewicz, E.M.; Lekka, M.

    1997-01-01

    Trace element analysis is one of the most important field in analytical chemistry. There are several instrumental techniques which are applied for determinations of microscopic elemental content. The PIXE (Proton Induced X-ray Emission) technique is one of the nuclear techniques that is commonly applied for such purpose due to its multielemental analysis possibilities. The aim of this study was to establish the optimal conditions for target preparation procedure. In this paper two different approaches to the topic are presented and widely discussed. The first approach was the traditional pellet technique and the second one was mineralization procedure. For the analysis soft tissue such as liver was used. Some results are also presented on water samples. (author)

  14. Aspects of cell calculations in deterministic reactor core analysis

    International Nuclear Information System (INIS)

    Varvayanni, M.; Savva, P.; Catsaros, N.

    2011-01-01

    Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

  15. Heating analysis of cobalt adjusters in reactor core

    International Nuclear Information System (INIS)

    Mei Qiliang; Li Kang; Fu Yaru

    2011-01-01

    In order to produce 60 Co source for industry and medicine applications in CANDU-6 reactor, the stainless steel adjusters were replaced with the cobalt adjusters. The cobalt rod will generate the heat when it is irradiated by neutron and γ ray. In addition, 59 Co will be activated and become 60 Co, the ray released due to 60 Co decay will be absorbed by adjusters, and then the adjusters will also generate the heat. So the heating rate of adjusters to be changed during normal operation must be studied, which will be provided as the input data for analyzing the temperature field of cobalt adjusters and the relative heat load of moderator. MCNP code was used to simulate whole core geometric configuration in detail, including reactor fuel, control rod, adjuster, coolant and moderator, and to analyze the heating rate of the stainless steel adjusters and the cobalt adjusters. The maximum heating rate of different cobalt adjuster based on above results will be provided for the steady thermal hydraulic and accident analysis, and make sure that the reactor is safe on the thermal hydraulic. (authors)

  16. Homogeneous protein analysis by magnetic core-shell nanorod probes

    KAUST Repository

    Schrittwieser, Stefan

    2016-03-29

    Studying protein interactions is of vital importance both to fundamental biology research and to medical applications. Here, we report on the experimental proof of a universally applicable label-free homogeneous platform for rapid protein analysis. It is based on optically detecting changes in the rotational dynamics of magnetically agitated core-shell nanorods upon their specific interaction with proteins. By adjusting the excitation frequency, we are able to optimize the measurement signal for each analyte protein size. In addition, due to the locking of the optical signal to the magnetic excitation frequency, background signals are suppressed, thus allowing exclusive studies of processes at the nanoprobe surface only. We study target proteins (soluble domain of the human epidermal growth factor receptor 2 - sHER2) specifically binding to antibodies (trastuzumab) immobilized on the surface of our nanoprobes and demonstrate direct deduction of their respective sizes. Additionally, we examine the dependence of our measurement signal on the concentration of the analyte protein, and deduce a minimally detectable sHER2 concentration of 440 pM. For our homogeneous measurement platform, good dispersion stability of the applied nanoprobes under physiological conditions is of vital importance. To that end, we support our measurement data by theoretical modeling of the total particle-particle interaction energies. The successful implementation of our platform offers scope for applications in biomarker-based diagnostics as well as for answering basic biology questions.

  17. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  18. Study of core support barrel vibration monitoring using ex-core neutron noise analysis and fuzzy logic algorithm

    International Nuclear Information System (INIS)

    Christian, Robby; Song, Seon Ho; Kang, Hyun Gook

    2015-01-01

    The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

  19. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  20. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    Rimpault, Gerald

    2003-01-01

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  1. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  2. Knowledge Economy Core Journals: Identification through LISTA Database Analysis.

    Science.gov (United States)

    Nouri, Rasool; Karimi, Saeed; Ashrafi-rizi, Hassan; Nouri, Azadeh

    2013-03-01

    Knowledge economy has become increasingly broad over the years and identification of core journals in this field can be useful for librarians in journal selection process and also for researchers to select their studies and finding Appropriate Journal for publishing their articles. Present research attempts to determine core journals of Knowledge Economy indexed in LISTA (Library and Information Science and Technology). The research method was bibliometric and research population include the journals indexed in LISTA (From the start until the beginning of 2011) with at least one article a bout "knowledge economy". For data collection, keywords about "knowledge economy"-were extracted from the literature in this area-have searched in LISTA by using title, keyword and abstract fields and also taking advantage of LISTA thesaurus. By using this search strategy, 1608 articles from 390 journals were retrieved. The retrieved records import in to the excel sheet and after that the journals were grouped and the Bradford's coefficient was measured for each group. Finally the average of the Bradford's coefficients were calculated and core journals with subject area of "Knowledge economy" were determined by using Bradford's formula. By using Bradford's scattering law, 15 journals with the highest publication rates were identified as "Knowledge economy" core journals indexed in LISTA. In this list "Library and Information update" with 64 articles was at the top. "ASLIB Proceedings" and "Serials" with 51 and 40 articles are next in rank. Also 41 journals were identified as beyond core that "Library Hi Tech" with 20 articles was at the top. Increased importance of knowledge economy has led to growth of production of articles in this subject area. So the evaluation of journals for ranking these journals becomes a very challenging task for librarians and generating core journal list can provide a useful tool for journal selection and also quick and easy access to information. Core

  3. CARVEDILOL POPULATION PHARMACOKINETIC ANALYSIS – APPLIED VALIDATION PROCEDURE

    Directory of Open Access Journals (Sweden)

    Aleksandra Catić-Đorđević

    2013-09-01

    Full Text Available Carvedilol is a nonselective beta blocker/alpha-1 blocker, which is used for treatment of essential hypertension, chronic stable angina, unstable angina and ischemic left ventricular dysfunction. The aim of this study was to describe carvedilol population pharmacokinetic (PK analysis as well as the validation of analytical procedure, which is an important step regarding this approach. In contemporary clinical practice, population PK analysis is often more important than standard PK approach in setting a mathematical model that describes the PK parameters. Also, it includes the variables that have particular importance in the drugs pharmacokinetics such as sex, body mass, dosage, pharmaceutical form, pathophysiological state, disease associated with the organism or the presence of a specific polymorphism in the isoenzyme important for biotransformation of the drug. One of the most frequently used approach in population PK analysis is the Nonlinear Modeling of Mixed Effects - NONMEM modeling. Analytical methods used in the data collection period is of great importance for the implementation of a population PK analysis of carvedilol in order to obtain reliable data that can be useful in clinical practice. High performance liquid chromatography (HPLC analysis of carvedilol is used to confirm the identity of a drug and provide quantitative results and also to monitor the efficacy of the therapy. Analytical procedures used in other studies could not be fully implemented in our research as it was necessary to perform certain modification and validation of the method with the aim of using the obtained results for the purpose of a population pharmacokinetic analysis. Validation process is a logical terminal phase of analytical procedure development that provides applicability of the procedure itself. The goal of validation is to ensure consistency of the method and accuracy of results or to confirm the selection of analytical method for a given sample

  4. The utilisation of thermal analysis to optimise radiocarbon dating procedures

    International Nuclear Information System (INIS)

    Brandova, D.; Keller, W.A.; Maciejewski, M.

    1999-01-01

    Thermal analysis combined with mass spectrometry was applied to radiocarbon dating procedures (age determination of carbon-containing samples). Experiments carried out under an oxygen atmosphere were used to determine carbon content and combustion range of soil and wood samples. Composition of the shell sample and its decomposition were investigated. The quantification of CO 2 formed by the oxidation of carbon was done by the application of pulse thermal analysis. Experiments carried out under an inert atmosphere determined the combustion range of coal with CuO as an oxygen source. To eliminate a possible source of contamination in the radiocarbon dating procedures the adsorption of CO 2 by CuO was investigated. (author)

  5. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

    2006-01-01

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  6. Structural analysis and optimization procedure of the TFTR device substructure

    International Nuclear Information System (INIS)

    Driesen, G.

    1975-10-01

    A structural evaluation of the TFTR device substructure is performed in order to verify the feasibility of the proposed design concept as well as to establish a design optimization procedure for minimizing the material and fabrication cost of the substructure members. A preliminary evaluation of the seismic capability is also presented. The design concept on which the analysis is based is consistent with that described in the Conceptual Design Status Briefing report dated June 18, 1975

  7. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  8. Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun

    2001-08-01

    This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

  9. Analysis of stress in reactor core vessel under effect of pressure lose shock wave

    International Nuclear Information System (INIS)

    Li Yong; Liu Baoting

    2001-01-01

    High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

  10. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  11. Analyses on the BFS critical experiments. An analysis on the BFS-62-1 and 62-2 cores

    International Nuclear Information System (INIS)

    Sugino, Kazuteru; Shono, Akira

    2002-04-01

    In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO 2 fuel surrounded by the UO 2 blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shied. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 cores analyses is

  12. Feasibility study of applying a multi-channel analysis model to on-line core monitoring system

    International Nuclear Information System (INIS)

    In, W. K.; Yoo, Y. J.; Hwang, D. H.; Jun, T. H.

    1998-01-01

    A feasibility study was performed to evaluate the effect of implementing a multi-channel analysis model in on-line core monitoring system. A simplified thermal-hydraulic model has been used in the on-line core monitoring system of digital PWR. The design procedure, core thermal margin and computation time were investigated in case of replacing the simplified model with the multi-channel analysis model. For the given ranges of limiting conditions for operation in Yonggwang Unit 3 Cycle 1, the minimum DNBR of the simplified thermal-hydraulic code CETOP-D was compared to that of the multi-channel analysis code MATRA. A CETOP-D tuning is additionally required to ensure the accurate and conservative DNBR calculation but the MATRA tuning is not necessary. MATRA appeared to increase the DNBR overpower margin from 2.5% to 6% over the CETOP-D margin. MATRA took approximately 1 second to compute DNBR on the HP9000 workstation system, which is longer than the DNBR computation time of CETOP-D. It is, however, fast enough to perform the on-line monitoring of DNBR. It can be therefore concluded that the application of the multi-channel analysis model MATRA in the on-line core monitoring system is feasible

  13. Design and analysis of three-layer-core optical fiber

    Science.gov (United States)

    Zheng, Siwen; Liu, Yazhuo; Chang, Guangjian

    2018-03-01

    A three-layer-core single-mode large-mode-area fiber is investigated. The three-layer structure in the core, which is composed of a core-index layer, a cladding-index layer, and a depression-index layer, could achieve a large effective area Aeff while maintaining an ultralow bending loss without deteriorating cutoff behaviors. The single-mode large mode area of 100 to 330 μm2 could be achieved in the fiber. The effective area Aeff can be further enlarged by adjusting the layer parameters. Furthermore, the bending property could be improved in this three-layer-core structure. The bending loss could decrease by 2 to 4 orders of magnitude compared with the conventional step-index fiber with the same Aeff. These characteristics of three-layer-core fiber suggest that it can be used in large-mode-area wide-bandwidth high-capacity transmission or high-power optical fiber laser and amplifier in optical communications, which could be used for the basic physical layer structure of big data storage, reading, calculation, and transmission applications.

  14. Analysis on First Criticality Benchmark Calculation of HTR-10 Core

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

    HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

  15. Two-dimensional horizontal model seismic test and analysis for HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1988-05-01

    The resistance against earthquakes of high-temperature gas-cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. The paper presented the test results of seismic behavior of a half scale two-dimensional horizontal slice core model and analysis. The following is a summary of the more important results. (1) When the core is subjected to the single axis excitation and simultaneous two-axis excitations to the core across-corners, it has elliptical motion. The core stays lumped motion at the low excitation frequencies. (2) When the load is placed on side fixed reflector blocks from outside to the core center, the core displacement and reflector impact reaction force decrease. (3) The maximum displacement occurs at simultaneous two-axis excitations. The maximum displacement occurs at the single axis excitation to the core across-flats. (4) The results of two-dimensional horizontal slice core model was compared with the results of two-dimensional vertical one. It is clarified that the seismic response of actual core can be predicted from the results of two-dimensional vertical slice core model. (5) The maximum reflector impact reaction force for seismic waves was below 60 percent of that for sinusoidal waves. (6) Vibration behavior and impact response are in good agreement between test and analysis. (author)

  16. A fast converging CFD model for thermal hydraulic analysis of gas cooled reactor cores

    International Nuclear Information System (INIS)

    Chen, Gary; Anghaie, Samim

    1999-01-01

    A computational fluid dynamics (CFD) approach to the solution of Navier-Stokes equations for the thermal and flow fields of gas cooled reactor cores is presented. An implicit-explicit MacCormack method based on finite volume discretization scheme, in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve axisymmetric, thin-layer Navier-Stokes equations. This numerical method requires only the inversion of block bidiagonal systems rather than block tridiagonal systems, thus yielding savings in computer time and storage requirements. A two-layer algebraic eddy viscosity turbulence model is used in this study. The effects of turbulence are simulated in terms of the eddy viscosity coefficient, which is calculated for an inner and an outer region separately. An enthalpy-rebalancing scheme is implemented to allow the convergence solutions to be obtained with the application of a wall heat flux. The detailed computational analysis developed in this work is used to evaluate many different Nusselt number equations, property corrections, and axial distance corrections. The calculation based on this CFD model is compared with other published results. The good agreement indicates the usefulness of the presented model for the prediction of flow and temperature distributions for gas cooled reactor cores. (author)

  17. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  18. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  19. Analysis of high moderation full MOX BWR core physics experiments BASALA

    International Nuclear Information System (INIS)

    Ishii, Kazuya; Ando, Yoshihira; Takada, Naoyuki; Kan, Taro; Sasagawa, Masaru; Kikuchi, Tsukasa; Yamamoto, Toru; Kanda, Ryoji; Umano, Takuya

    2005-01-01

    Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO 2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO 2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores. (author)

  20. Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

    1995-08-01

    The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

  1. Effect analysis of core barrel openings under CEFR normal condition

    International Nuclear Information System (INIS)

    Zhang Yabo; Yang Hongyi

    2008-01-01

    Openings on the bottom of core barrel are important part of the decay heat removal system of China Experimental Fast Reactor (CEFR), which are designed to discharge the decay heat from reactor under accident condition. This paper analyses the effect of the openings design on the normal operation condition using the famouse CFD code CFX. The result indicates that the decay heat can be discharged safely and at the same time the effect of core barrel openings on the normal operation condition is acceptable. (authors)

  2. Extraction of trapped gases in ice cores for isotope analysis

    International Nuclear Information System (INIS)

    Leuenberger, M.; Bourg, C.; Francey, R.; Wahlen, M.

    2002-01-01

    The use of ice cores for paleoclimatic investigations is discussed in terms of their application for dating, temperature indication, spatial time marker synchronization, trace gas fluxes, solar variability indication and changes in the Dole effect. The different existing techniques for the extraction of gases from ice cores are discussed. These techniques, all to be carried out under vacuum, are melt-extraction, dry-extraction methods and the sublimation technique. Advantages and disadvantages of the individual methods are listed. An extensive list of references is provided for further detailed information. (author)

  3. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    improvements address both neutronics and thermal-hydraulics aspects. Furthermore, emphasis has been placed on not only the beginning-of-life (BOL) state of the core, but also on the beginning of closed equilibrium fuel cycle (BEC) state. An important context for the current thesis is the 7{sup th} European Framework Program's Collaborative Project for a European Sodium Fast Reactor (CP-ESFR), the reference 3600 MWth ESFR core being the starting point for the conducted research. The principally employed computational tools belong to the so-called FAST code system, viz. the fast-reactor neutronics code ERANOS, the fuel cycle simulating procedure EQL3D, the spatial kinetics code PARCS and the system thermal-hydraulics code TRACE. The research has been carried out in essentially three successive phases. The first phase has involved achieving a clearer understanding of the principal phenomena contributing to the SFR void effect. Decomposition and analysis of sodium void reactivity have been carried out, while considering different fuel cycle states for the core. Furthermore, the spatial distribution of void reactivity importance, in both axial and radial directions, is investigated. For the reactivity decomposition, two methods, based respectively on neutron balance considerations and on perturbation theory, have been applied. The sodium void reactivity of the reference ESFR core has been, accordingly, decomposed reaction-wise, cross-section-wise, isotope-wise and energy-group-wise. Effectively, the neutron balance based method allows an in-depth understanding of the ‘consequences’ of sodium voidage, while the perturbation theory based method provides a complementary understanding of the ‘causes’. The second phase of the research has addressed optimization of the reference ESFR core design from the neutronics viewpoint. Four options oriented towards either the leakage component or the spectral effect have been considered in detail, viz. introducing an upper sodium

  4. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    improvements address both neutronics and thermal-hydraulics aspects. Furthermore, emphasis has been placed on not only the beginning-of-life (BOL) state of the core, but also on the beginning of closed equilibrium fuel cycle (BEC) state. An important context for the current thesis is the 7 th European Framework Program's Collaborative Project for a European Sodium Fast Reactor (CP-ESFR), the reference 3600 MWth ESFR core being the starting point for the conducted research. The principally employed computational tools belong to the so-called FAST code system, viz. the fast-reactor neutronics code ERANOS, the fuel cycle simulating procedure EQL3D, the spatial kinetics code PARCS and the system thermal-hydraulics code TRACE. The research has been carried out in essentially three successive phases. The first phase has involved achieving a clearer understanding of the principal phenomena contributing to the SFR void effect. Decomposition and analysis of sodium void reactivity have been carried out, while considering different fuel cycle states for the core. Furthermore, the spatial distribution of void reactivity importance, in both axial and radial directions, is investigated. For the reactivity decomposition, two methods, based respectively on neutron balance considerations and on perturbation theory, have been applied. The sodium void reactivity of the reference ESFR core has been, accordingly, decomposed reaction-wise, cross-section-wise, isotope-wise and energy-group-wise. Effectively, the neutron balance based method allows an in-depth understanding of the ‘consequences’ of sodium voidage, while the perturbation theory based method provides a complementary understanding of the ‘causes’. The second phase of the research has addressed optimization of the reference ESFR core design from the neutronics viewpoint. Four options oriented towards either the leakage component or the spectral effect have been considered in detail, viz. introducing an upper sodium plenum

  5. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  6. Analysis of Elementary School students’ algebraic perceptions and procedures

    Directory of Open Access Journals (Sweden)

    Sandra Mara Marasini

    2012-12-01

    Full Text Available This study aims to verify how students in elementary school see themselves in relation to mathematics and, at the same time, analyze the procedures used to solve algebraic tasks. These students in the 8th year of elementary school, and first and third years of high school, from two State schools in Passo Fundo/RS, answered a questionnaire about their own perceptions of the mathematics lessons, the subject mathematics and algebraic content. The analysis was based mainly on authors from the athematical education and the historic-cultural psychology areas. It was verifi ed that even among students who claimed to be happy with the idea of having mathematicsclasses several presented learning diffi culties regarding algebraic contents, revealed by the procedures employed. It was concluded that it is necessary to design proposals with didactic sequences, mathematically and pedagogically based, which can effi cientlyoptimize the appropriation of meaning from the concepts approached and their application in different situations.

  7. Safety analysis of the topaz behavior during irradiation, its effect on the core performance and the in-core fuel management strategy

    International Nuclear Information System (INIS)

    Khalil, M.Y.; Belal, M.G.

    2006-01-01

    The topaz is a natural gem stones which collect color centers when irradiated with fast neutrons and transformed into a colorful stones called topaz. The objective of this paper is to detail the safety analysis performed to assure the safety measures of the topaz mass production and farther shows an indirect estimated measurement of the safety related parameters. Analysis has been performed for all the irradiation positions nominated for topaz production and this paper present experimental verification performed for the position of the highest influence where all other positions have lower influences and showed the same safety features and agreement between calculations and measurements. On the other hand it was necessary to show that no hot spots and no cooling problems would rise as a result of irradiation. The heat energy dissipation in the topaz boxes is important from the reactor core coolability side as well as from the view point of the quality of the product. Moreover the paper describes the administrative procedure to limit the reactivity insertion rate of any box to less than 10 pcm/sec. The effect of the topaz boxes presence on the accumulated fuel burn up has been calculated, and recommendations concerning the in-core fuel management strategy has been reviewed. (authors)

  8. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

  9. Application of a statistical thermal design procedure to evaluate the PWR DNBR safety analysis limits

    International Nuclear Information System (INIS)

    Robeyns, J.; Parmentier, F.; Peeters, G.

    2001-01-01

    In the framework of safety analysis for the Belgian nuclear power plants and for the reload compatibility studies, Tractebel Energy Engineering (TEE) has developed, to define a 95/95 DNBR criterion, a statistical thermal design method based on the analytical full statistical approach: the Statistical Thermal Design Procedure (STDP). In that methodology, each DNBR value in the core assemblies is calculated with an adapted CHF (Critical Heat Flux) correlation implemented in the sub-channel code Cobra for core thermal hydraulic analysis. The uncertainties of the correlation are represented by the statistical parameters calculated from an experimental database. The main objective of a sub-channel analysis is to prove that in all class 1 and class 2 situations, the minimum DNBR (Departure from Nucleate Boiling Ratio) remains higher than the Safety Analysis Limit (SAL). The SAL value is calculated from the Statistical Design Limit (SDL) value adjusted with some penalties and deterministic factors. The search of a realistic value for the SDL is the objective of the statistical thermal design methods. In this report, we apply a full statistical approach to define the DNBR criterion or SDL (Statistical Design Limit) with the strict observance of the design criteria defined in the Standard Review Plan. The same statistical approach is used to define the expected number of rods experiencing DNB. (author)

  10. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    Schulz, K.C.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K Q due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  11. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Mohamed, F.M.

    2011-01-01

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  12. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    International Nuclear Information System (INIS)

    Fridman, E.; Kliem, S.

    2011-01-01

    Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

  13. Analysis of a multigroup stylized CANDU half-core benchmark

    International Nuclear Information System (INIS)

    Pounders, Justin M.; Rahnema, Farzad; Serghiuta, Dumitru

    2011-01-01

    Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

  14. 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.

  15. Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

    International Nuclear Information System (INIS)

    Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

    2012-01-01

    The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

  16. Beacon: A three-dimensional structural analysis code for bowing history of fast breeder reactor cores

    International Nuclear Information System (INIS)

    Miki, K.

    1979-01-01

    The core elements of an LMFBR are bowed due to radial gradients of both temperature and neutron flux in the core. Since all hexagonal elements are multiply supported by adjacent elements or the restraint system, restraint forces and bending stresses are induced. In turn, these forces and stresses are relaxed by irradiation enhanced creep of the material. The analysis of the core bowing behavior requires a three-dimensional consideration of the mechanical interactions among the core elements, because the core consists of different kinds of elements and of fuel assemblies with various burnup histories. A new computational code BEACON has been developed for analyzing the bowing behavior of an LMFBR's core in three dimensions. To evaluate mechanical interactions among core elements, the code uses the analytical method of the earlier SHADOW code. BEACON analyzes the mechanical interactions in three directions, which form angles of 60 0 with one another. BEACON is applied to the 60 0 sector of a typical LMFBR's core for analyzing the bowing history during one equilibrium cycle. 120 core elements are treated, assuming the boundary condition of rotational symmetry. The application confirms that the code can be an effective tool for parametric studies as well as for detailed structural analysis of LMFBR's core. (orig.)

  17. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  18. Improvement of Axial Reflector Cross Section Generation Model for PWR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Cheon Bo; Lee, Kyung Hoon; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper covers the study for improvement of axial reflector XS generation model. In the next section, the improved 1D core model is represented in detail. Reflector XS generated by the improved model is compared to that of the conventional model in the third section. Nuclear design parameters generated by these two XS sets are also covered in that section. Significant of this study is discussed in the last section. Two-step procedure has been regarded as the most practical approach for reactor core designs because it offers core design parameters quite rapidly within acceptable range. Thus this approach is adopted for SMART (System-integrated Modular Advanced Reac- Tor) core design in KAERI with the DeCART2D1.1/ MASTER4.0 (hereafter noted as DeCART2D/ MASTER) code system. Within the framework of the two-step procedure based SMART core design, various researches have been studied to improve the core design reliability and efficiency. One of them is improvement of reflector cross section (XS) generation models. While the conventional FA/reflector two-node model used for most core designs to generate reflector XS cannot consider the actual configuration of fuel rods that intersect at right angles to axial reflectors, the revised model reflects the axial fuel configuration by introducing the radially simplified core model. The significance of the model revision is evaluated by observing HGC generated by DeCART2D, reflector XS, and core design parameters generated by adopting the two models. And it is verified that about 30 ppm CBC error can be reduced and maximum Fq error decreases from about 6 % to 2.5 % by applying the revised model. Error of AO and axial power shapes are also reduced significantly. Therefore it can be concluded that the simplified 1D core model improves the accuracy of the axial reflector XS and leads to the two-step procedure reliability enhancement. Since it is hard for core designs to be free from the two-step approach, it is necessary to find

  19. Elemental hair analysis: A review of procedures and applications

    International Nuclear Information System (INIS)

    Pozebon, D.; Scheffler, G.L.; Dressler, V.L.

    2017-01-01

    Although exogenous contamination and unreliable reference values have limited the utility of scalp hair as a biomarker of chemical elements exposure, its use in toxicological, clinical, environmental and forensic investigations is growing and becoming more extensive. Therefore, hair elemental analysis is reviewed in the current manuscript which spans articles published in the last 10 years. It starts with a general discussion of history, morphology and possible techniques for elemental analysis, where inductively coupled plasma-mass spectrometry (ICP-MS) is clearly highlighted since this technique is leading quantitative ultra-trace elemental analysis. Emphasis over sampling, quality assurance, washing procedures and sample decomposition is given with detailed protocols compiled in tables as well as the utility of hair to identify human gender, age, diseases, healthy conditions, nutrition status and contamination sites. Isotope ratio information, chemical speciation analysis and analyte preconcentration are also considered for hair. Finally, the potential of laser ablation ICP-MS (LA-ICP-MS) to provide spatial resolution and time-track the monitoring of elements in hair strands instead of conventional bulk analysis is spotlighted as a real future trend in the field. - Highlights: • Elemental analysis of hair is critically reviewed, with focus on ICP-MS employment. • Standards protocols of hair washing and sample decomposition are compiled. • The usefulness of elemental and/or isotopic analysis of hair is demonstrated. • The potential of LA-ICP-MS for elemental time tracking in hair is highlighted.

  20. A simplified procedure of linear regression in a preliminary analysis

    Directory of Open Access Journals (Sweden)

    Silvia Facchinetti

    2013-05-01

    Full Text Available The analysis of a statistical large data-set can be led by the study of a particularly interesting variable Y – regressed – and an explicative variable X, chosen among the remained variables, conjointly observed. The study gives a simplified procedure to obtain the functional link of the variables y=y(x by a partition of the data-set into m subsets, in which the observations are synthesized by location indices (mean or median of X and Y. Polynomial models for y(x of order r are considered to verify the characteristics of the given procedure, in particular we assume r= 1 and 2. The distributions of the parameter estimators are obtained by simulation, when the fitting is done for m= r + 1. Comparisons of the results, in terms of distribution and efficiency, are made with the results obtained by the ordinary least square methods. The study also gives some considerations on the consistency of the estimated parameters obtained by the given procedure.

  1. Reduction procedures for accurate analysis of MSX surveillance experiment data

    Science.gov (United States)

    Gaposchkin, E. Mike; Lane, Mark T.; Abbot, Rick I.

    1994-01-01

    Technical challenges of the Midcourse Space Experiment (MSX) science instruments require careful characterization and calibration of these sensors for analysis of surveillance experiment data. Procedures for reduction of Resident Space Object (RSO) detections will be presented which include refinement and calibration of the metric and radiometric (and photometric) data and calculation of a precise MSX ephemeris. Examples will be given which support the reduction, and these are taken from ground-test data similar in characteristics to the MSX sensors and from the IRAS satellite RSO detections. Examples to demonstrate the calculation of a precise ephemeris will be provided from satellites in similar orbits which are equipped with S-band transponders.

  2. Automated procedure for performing computer security risk analysis

    International Nuclear Information System (INIS)

    Smith, S.T.; Lim, J.J.

    1984-05-01

    Computers, the invisible backbone of nuclear safeguards, monitor and control plant operations and support many materials accounting systems. Our automated procedure to assess computer security effectiveness differs from traditional risk analysis methods. The system is modeled as an interactive questionnaire, fully automated on a portable microcomputer. A set of modular event trees links the questionnaire to the risk assessment. Qualitative scores are obtained for target vulnerability, and qualitative impact measures are evaluated for a spectrum of threat-target pairs. These are then combined by a linguistic algebra to provide an accurate and meaningful risk measure. 12 references, 7 figures

  3. Review and Application of Ship Collision and Grounding Analysis Procedures

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    2010-01-01

    It is the purpose of the paper to present a review of prediction and analysis tools for collision and grounding analyses and to outline a probabilistic procedure for which these tools can be used by the maritime industry to develop performance based rules to reduce the risk associated with human,......, environmental and economic costs of collision and grounding events. The main goal of collision and grounding research should be to identify the most economic risk control options associated with prevention and mitigation of collision and grounding events....

  4. User's operating procedures. Volume 2: Scout project financial analysis program

    Science.gov (United States)

    Harris, C. G.; Haris, D. K.

    1985-01-01

    A review is presented of the user's operating procedures for the Scout Project Automatic Data system, called SPADS. SPADS is the result of the past seven years of software development on a Prime mini-computer located at the Scout Project Office, NASA Langley Research Center, Hampton, Virginia. SPADS was developed as a single entry, multiple cross-reference data management and information retrieval system for the automation of Project office tasks, including engineering, financial, managerial, and clerical support. This volume, two (2) of three (3), provides the instructions to operate the Scout Project Financial Analysis program in data retrieval and file maintenance via the user friendly menu drivers.

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

    International Nuclear Information System (INIS)

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

    2002-04-01

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

  6. Magnetic loss analysis in Mn-Zn ferrite cores

    International Nuclear Information System (INIS)

    Beatrice, C.; Bottauscio, O.; Chiampi, M.; Fiorillo, F.; Manzin, A.

    2006-01-01

    Magnetic losses have been measured and analyzed upon a wide range of frequencies in Mn-Zn ferrite ring cores. Exploiting the concept of loss separation and modeling the conductivity process in the heterogeneous material as a function of frequency, the role of the different energy dissipation mechanisms has been elucidated. It is shown, in particular, that eddy current effects can be appreciated, in standard materials and cores, only on approaching and overcoming the MHz range. The basic mechanism for hysteresis and low-frequency losses is therefore identified with the domain wall relaxation engendered by spin damping processes. Resonant absorption of energy associated with magnetization rotation is in turn deemed to chiefly contribute to the loss upon the practical range of frequencies going from a few 10 4 Hz to a few MHz

  7. Analysis of a ferrofluid core differential transformer tilt measurement sensor

    Energy Technology Data Exchange (ETDEWEB)

    Medvegy, T.; Molnár, Á.; Molnár, G.; Gugolya, Z.

    2017-04-15

    In our work, we developed a ferrofluid core differential transformer sensor, which can be used to measure tilt and acceleration. The proposed sensor consisted of three coils, from which the primary was excited with an alternating current. In the space surrounded by the coils was a cell half-filled with ferrofluid, therefore in the horizontal state of the sensor the fluid distributes equally in the three sections of the cell surrounded by the three coils. Nevertheless when the cell is being tilted or accelerated (in the direction of the axis of the coils), there is a different amount of ferrofluid in the three sections. The voltage induced in the secondary coils strongly depends on the amount of ferrofluid found in the core surrounded by them, so the tilt or the acceleration of the cell becomes measurable. We constructed the sensor in several layouts. The linearly coiled sensor had an excellent resolution. Another version with a toroidal cell had almost perfect linearity and a virtually infinite measuring range. - Highlights: • A ferrofluid core differential transformer can be used to measure tilt. • The theoretical description of two different type of the sensor is introduced. • The measuring range, and the sensitivity depends on the dimensions of the sensor.

  8. Analysis of core samples from jet grouted soil

    International Nuclear Information System (INIS)

    Allan, M.L.; Kukacka, L.E.

    1995-10-01

    Superplasticized cementitious grouts were tested for constructing subsurface containment barriers using jet grouting in July, 1994. The grouts were developed in the Department of Applied Science at Brookhaven National Laboratory. The test site was located close to the Chemical Waste Landfill at Sandia National Laboratories, Albuquerque, NM. Sandia was responsible for the placement contract. The jet grouted soil was exposed to the service environment for one year and core samples were extracted to evaluate selected properties. The cores were tested for strength, density, permeability (hydraulic conductivity) and cementitious content. The tests provided an opportunity to determine the performance of the grouts and grout-treated soil. Several recommendations arise from the results of the core tests. These are: (1) grout of the same mix proportions as the final grout should be used as a drilling fluid in order to preserve the original mix design and utilize the benefits of superplasticizers; (2) a high shear mixer should be used for preparation of the grout; (3) the permeability under unsaturated conditions requires consideration when subsurface barriers are used in the vadose zone; and (4) suitable methods for characterizing the permeability of barriers in-situ should be applied

  9. Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

    International Nuclear Information System (INIS)

    Nakagawa, M.; Tsuboi, Y.

    1990-01-01

    ''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

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

    International Nuclear Information System (INIS)

    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

  11. Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

    International Nuclear Information System (INIS)

    Gehre, G.

    1982-01-01

    First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

  12. Deconvolution-based resolution enhancement of chemical ice core records obtained by continuous flow analysis

    DEFF Research Database (Denmark)

    Rasmussen, Sune Olander; Andersen, Katrine K.; Johnsen, Sigfus Johann

    2005-01-01

    Continuous flow analysis (CFA) has become a popular measuring technique for obtaining high-resolution chemical ice core records due to an attractive combination of measuring speed and resolution. However, when analyzing the deeper sections of ice cores or cores from low-accumulation areas...... of the data for high-resolution studies such as annual layer counting. The presented method uses deconvolution techniques and is robust to the presence of noise in the measurements. If integrated into the data processing, it requires no additional data collection. The method is applied to selected ice core...

  13. Shrunken head (tsantsa): a complete forensic analysis procedure.

    Science.gov (United States)

    Charlier, P; Huynh-Charlier, I; Brun, L; Hervé, C; de la Grandmaison, G Lorin

    2012-10-10

    Based on the analysis of shrunken heads referred to our forensic laboratory for anthropological expertise, and data from both anthropological and medical literature, we propose a complete forensic procedure for the analysis of such pieces. A list of 14 original morphological criteria has been developed, based on the global aspect, color, physical deformation, anatomical details, and eventual associated material (wood, vegetal fibers, sand, charcoals, etc.). Such criteria have been tested on a control sample of 20 tsantsa (i.e. shrunken heads from the Jivaro or Shuar tribes of South America). Further complementary analyses are described such as CT-scan and microscopic examination. Such expertise is more and more asked to forensic anthropologists and practitioners in a context of global repatriation of human artifacts to native communities. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.

  14. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  15. Performance modeling and analysis of parallel Gaussian elimination on multi-core computers

    Directory of Open Access Journals (Sweden)

    Fadi N. Sibai

    2014-01-01

    Full Text Available Gaussian elimination is used in many applications and in particular in the solution of systems of linear equations. This paper presents mathematical performance models and analysis of four parallel Gaussian Elimination methods (precisely the Original method and the new Meet in the Middle –MiM– algorithms and their variants with SIMD vectorization on multi-core systems. Analytical performance models of the four methods are formulated and presented followed by evaluations of these models with modern multi-core systems’ operation latencies. Our results reveal that the four methods generally exhibit good performance scaling with increasing matrix size and number of cores. SIMD vectorization only makes a large difference in performance for low number of cores. For a large matrix size (n ⩾ 16 K, the performance difference between the MiM and Original methods falls from 16× with four cores to 4× with 16 K cores. The efficiencies of all four methods are low with 1 K cores or more stressing a major problem of multi-core systems where the network-on-chip and memory latencies are too high in relation to basic arithmetic operations. Thus Gaussian Elimination can greatly benefit from the resources of multi-core systems, but higher performance gains can be achieved if multi-core systems can be designed with lower memory operation, synchronization, and interconnect communication latencies, requirements of utmost importance and challenge in the exascale computing age.

  16. Pertinent anatomy and analysis for midface volumizing procedures.

    Science.gov (United States)

    Surek, Christopher C; Beut, Javier; Stephens, Robert; Jelks, Glenn; Lamb, Jerome

    2015-05-01

    The study was conducted to construct an anatomically inspired midfacial analysis facilitating safe, accurate, and dynamic nonsurgical rejuvenation. Emphasis is placed on determining injection target areas and adverse event zones. Twelve hemifacial fresh cadavers were dissected in a layered fashion. Dimensional measurements between the midfacial fat compartments, prezygomatic space, mimetic muscles, and neurovascular bundles were used to develop a topographic analysis for clinical injections. A longitudinal line from the base of the alar crease to the medial edge of the levator anguli oris muscle (1.9 cm), lateral edge of the levator anguli oris muscle (2.6 cm), and zygomaticus major muscle (4.6 cm) partitions the cheek into two aesthetic regions. A six-step facial analysis outlines three target zones and two adverse event zones and triangulates the point of maximum cheek projection. The lower adverse event zone yields an anatomical explanation to inadvertent jowling during anterior cheek injection. The upper adverse event zone localizes the palpebral branch of the infraorbital artery. The medial malar target area isolates quadrants for anterior cheek projection and tear trough effacement. The middle malar target area addresses lid-cheek blending and superficial compartment turgor. The lateral malar target area highlights lateral cheek projection and locates the prezygomatic space. This stepwise analysis illustrates target areas and adverse event zones to achieve midfacial support, contour, and profile in the repose position and simultaneous molding of a natural shape during animation. This reproducible method can be used both procedurally and in record-keeping for midface volumizing procedures.

  17. A macroscopic cross-section model for BWR pin-by-pin core analysis

    International Nuclear Information System (INIS)

    Fujita, Tatsuya; Endo, Tomohiro; Yamamoto, Akio

    2014-01-01

    A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin core analysis is studied. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of core state and depletion history variables and are tabulated prior to core calculations. Variations of cross sections in a core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the core state variables and the exposure-averaged core state variables, respectively. Furthermore, the cross-term effect among the core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history (or the void history) and other variables have large impacts; thus, the consideration of them is crucial. The present macroscopic cross-section model, which considers such dominant cross-term effects, well reproduces the reference results and can be a candidate in practical applications for BWR pin-by-pin core analysis on the normal operations. (author)

  18. ARKAS: A three-dimensional finite element code for the analysis of core distortions and mechanical behaviour

    International Nuclear Information System (INIS)

    Nakagawa, M.

    1984-01-01

    Computer program ARKAS has been developed for the purpose of predicting core distortions and mechanical behaviour in a cluster of subassemblies under steady state conditions in LMFBR cores. This report describes the analytical models and numerical procedures employed in the code together with some typical results of the analysis made on large LMFBR cores. ARKAS is programmed in the FORTRAN-IV language and is capable of treating up to 260 assemblies in a cluster with flexible boundary conditions including mirror and rotational symmetry. The nonlinearity of the problem due to contact and separation is solved by the step iterative procedure based on the Newton-Raphson method. In each step iterative procedure, the linear matrix equation must be reconstructed and then solved directly. To save computer time and memory, the substructure method is adopted in the step of reconstructing the linear matrix equation, and in the step of solving the linear matrix equation, the block successive over-relaxation method is adopted. The program ARKAS computes, at every time step, 3-dimensional displacements and rotations of the subassemblies in the core and the interduct forces including at the nozzle tips and nozzle bases with friction effects. The code also has an ability to deal with the refueling and shuffling of subassemblies and to calculate the values of withdrawal forces. For the qualitative validation of the code, sample calculations were performed on the several bundle arrays. In these calculations, contact and separation processes under the influences of friction forces, off-center loading, duct rotations and torsion, thermal expansion and irradiation induced swelling and creep were analyzed. These results are quite reasonable in the light of the expected behaviour. This work was performed under the sponsorship of Toshiba Corporation

  19. Full core reactor analysis: Running Denovo on Jaguar

    Energy Technology Data Exchange (ETDEWEB)

    Jarrell, J. J.; Godfrey, A. T.; Evans, T. M.; Davidson, G. G. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2012-07-01

    Fully-consistent, full-core, 3D, deterministic neutron transport simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to more than 160 k processors. Cell-homogenized cross sections were used with step-characteristics, linear-discontinuous finite element, and trilinear-discontinuous finite element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large-aspect ratio meshes than using step-characteristics. (authors)

  20. Whole-core analysis by 13C NMR

    International Nuclear Information System (INIS)

    Vinegar, H.J.; Tutunjian, P.N.; Edelstein, W.A.; Roemer, P.B.

    1991-01-01

    This paper reports on a whole-core nuclear magnetic resonance (NMR) system that was used to obtain natural abundance 13 C spectra. The system enables rapid, nondestructive measurements of bulk volume of movable oil, aliphatic/aromatic ratio, oil viscosity, and organic vs. carbonate carbon. 13 C NMR can be used in cores where the 1 H NMR spectrum is too broad to resolve oil and water resonances separately. A 5 1/4-in. 13 C/ 1 H NMR coil was installed on a General Electric (GE) CSI-2T NMR imager/spectrometer. With a 4-in.-OD whole core, good 13 C signal/noise ratio (SNR) is obtained within minutes, while 1 H spectra are obtained in seconds. NMR measurements have been made of the 13 C and 1 H density of crude oils with a wide range of API gravities. For light- and medium-gravity oils, the 13 C and 1 H signal per unit volume is constant within about 3.5%. For heavy crudes, the 13 C and 1 H density measured by NMR is reduced by the shortening of spin-spin relaxation time. 13 C and 1 H NMR spin-lattice relaxation times were measured on a suite of Cannon viscosity standards, crude oils (4 to 60 degrees API), and alkanes (C 5 through C 16 ) with viscosities at 77 degrees F ranging from 0.5 cp to 2.5 x 10 7 cp. The 13 C and 1 H relaxation times show a similar correlation with viscosity from which oil viscosity can be estimated accurately for viscosities up to 100 cp. The 13 C surface relaxation rate for oils on water-wet rocks is very low. Nonproton decoupled 13 C NMR is shown to be insensitive to kerogen; thus, 13 C NMR measures only the movable hydrocarbon content of the cores. In carbonates, the 13 C spectrum also contains a carbonate powder pattern useful in quantifying inorganic carbon and distinguishing organic from carbonate carbon

  1. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

    1989-10-01

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  2. Comparative analysis of diagnostic accuracy of different brain biopsy procedures.

    Science.gov (United States)

    Jain, Deepali; Sharma, Mehar Chand; Sarkar, Chitra; Gupta, Deepak; Singh, Manmohan; Mahapatra, A K

    2006-12-01

    Image-guided procedures such as computed tomography (CT) guided, neuronavigator-guided and ultrasound-guided methods can assist neurosurgeons in localizing the intraparenchymal lesion of the brain. However, despite improvements in the imaging techniques, an accurate diagnosis of intrinsic lesion requires tissue sampling and histological verification. The present study was carried out to examine the reliability of the diagnoses made on tumor sample obtained via different stereotactic and ultrasound-guided brain biopsy procedures. A retrospective analysis was conducted of all brain biopsies (frame-based and frameless stereotactic and ultrasound-guided) performed in a single tertiary care neurosciences center between 1995 and 2005. The overall diagnostic accuracy achieved on histopathology and correlation with type of biopsy technique was evaluated. A total of 130 cases were included, which consisted of 82 males and 48 females. Age ranged from 4 to 75 years (mean age 39.5 years). Twenty per cent (27 patients) were in the pediatric age group, while 12% (16 patients) were >or= 60-years of age. A definitive histological diagnosis was established in 109 cases (diagnostic yield 80.2%), which encompassed 101 neoplastic and eight nonneoplastic lesions. Frame-based, frameless stereotactic and ultrasound-guided biopsies were done in 95, 15 and 20 patients respectively. Although the numbers of cases were small there was trend for better yield with frameless image-guided stereotactic biopsy and maximum diagnostic yield was obtained i.e, 87% (13/15) in comparison to conventional frame-based CT-guided stereotactic biopsy and ultrasound-guided biopsy. Overall, a trend of higher diagnostic yield was seen in cases with frameless image-guided stereotactic biopsy. Thus, this small series confirms that frameless neuronavigator-guided stereotactic procedures represent the lesion sufficiently in order to make histopathologic diagnosis.

  3. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    Dubuisson, B.; Smolarz, A.

    1984-01-01

    The study shows, with the results obtained, how it is possible to approach the problem of diagnosis with a technique based on the use of algorithms for statistical pattern recognition was justifiable. The results presented here, with a view to their use for fast breeder reactor core surveillance, are very encouraging, the most important point being the data representation. For this study, it was difficult to find the most suitable parameters for characterizing the various simulated core states, however, despite this handicap, the classification algorithm provided quite acceptable results. The second point concerns the characterization of a system's evolution. The criterion defined was chosen for adaptation to our algorithm. One acertained that it was possible to characterize evolution on the basis of this criterion as long as the rejected points were not too far from the known learning sets. Under these circumstances, the advantage in characterizing evolution in that the changes in evolution occur when the rejected points have a tendency to agglomerate in a small area of space could be seen. This phenomenon thus makes it possible to forsee whether the creation of a new class is possible. Where the rejected points are far away from the known learning sets, the criterion used proved to be too sensitive and the characterization of evolution was less satisfactory

  4. Analysis of core plasma heating and ignition by relativistic electrons

    International Nuclear Information System (INIS)

    Nakao, Y.

    2002-01-01

    Clarification of the pre-compressed plasma heating by fast electrons produced by relativistic laser-plasma interaction is one of the most important issues of the fast ignition scheme in ICF. On the basis of overall calculations including the heating process, both by relativistic hot electrons and alpha-particles, and the hydrodynamic evolution of bulk plasma, we examine the feature of core plasma heating and the possibility of ignition. The deposition of the electron energy via long-range collective mode, i.e. Langmuir wave excitation, is shown to be comparable to that through binary electron-electron collisions; the calculation neglecting the wave excitation considerably underestimates the core plasma heating. The ignition condition is also shown in terms of the intensity I(h) and temperature T(h) of hot electrons. It is found that I(h) required for ignition increases in proportion to T(h). For efficiently achieving the fast ignition, electron beams with relatively 'low' energy (e.g.T(h) below 1 MeV) are desirable. (author)

  5. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Rimpault, G.

    2004-01-01

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  6. 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.

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

    International Nuclear Information System (INIS)

    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

  8. Tank 241-B-203 push mode core sampling and analysis plan. Revision 1

    International Nuclear Information System (INIS)

    Jo, J.

    1995-01-01

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two push-mode core samples from tank 241-B-203 (B-203)

  9. Tank 241-B-204 push mode core sampling and analysis plan. Revision 1

    International Nuclear Information System (INIS)

    Sasaki, L.M.

    1995-01-01

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two push-mode core samples from tank 241-B-204 (B-204)

  10. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  11. An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

    International Nuclear Information System (INIS)

    Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

    2006-01-01

    KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

  12. Tank 241-U-105 push mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    Bell, K.E.

    1995-01-01

    This Sampling and Analysis Plan (SAP) will identify characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for vapor samples and two push mode core samples from tank 241-U-105 (U-105)

  13. Report on nuclear industry quality assurance procedures for safety analysis computer code development and use

    International Nuclear Information System (INIS)

    Sheron, B.W.; Rosztoczy, Z.R.

    1980-08-01

    As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement

  14. CT-guided core biopsy and percutaneous fiducial seed placement in the lung: Can these procedures be combined without an increase in complication rate or decrease in technical success?

    Energy Technology Data Exchange (ETDEWEB)

    Mendiratta-Lala, Mishal [Henry Ford Hospital, Department of Radiology, Abdominal Interventional Radiology, 2799 West Grand Blvd, Detroit, MI 48202 (United States); Sheiman, Robert, E-mail: rsheiman@bidmc.harvard.edu [Beth Israel Deaconess Hospital, Department of Radiology, Abdominal Imaging, One Deaconess Road, Boston, MA 02215 (United States); Brook, Olga R. [Beth Israel Deaconess Hospital, Department of Radiology, Abdominal Imaging, One Deaconess Road, Boston, MA 02215 (United States); Gourtsoyianni, Sofia [King' s College London, St Thomas’ Hospital, Lambeth Palace Road, SE1 7EH London (United Kingdom); Mahadevan, Anand [Beth Israel Deaconess Hospital, Radiation Oncology, One Deaconess Road, Boston, MA 02215 (United States); Siewert, Bettina [Beth Israel Deaconess Hospital, Department of Radiology, Abdominal Imaging, One Deaconess Road, Boston, MA 02215 (United States)

    2014-04-15

    Objective: To determine if concomitant CT-guided biopsy and percutaneous fiducial seed placement in the lung can be performed in a selective patient population without increased complication or decreased success rates compared to either procedure alone. Materials and methods: An IRB approved retrospective analysis of 285 consecutive patients that underwent CT-guided placement of fiducial seeds in the lung alone (N = 63), with concomitant core biopsy (N = 53) or only core biopsy (N = 169) was performed. Variables compared included: patient demographics, lesion size, depth from pleura, needle size, number of passes through pleura, number and size of core biopsies, number of seeds placed and technical success rates. Statistical analysis was performed using univariate and multivariate pair-wise comparisons. Results: A pathologic diagnosis of malignancy was confirmed in all cases undergoing seed placement alone and seed placement with concurrent biopsy, and in 144 of the biopsy alone lesions. On univariate analysis, major complication rates were similar for all three groups as were lesion size, depth, number of pleural passes, and technical success. Pair-wise comparisons of the remaining variables demonstrated a significant younger age and smaller needle size in the biopsy only group, and less minor complications in the fiducial only group. Overall there were 80/285 (28.1%) minor and 29/285 (10.2%) major complications. All major complications leading to admission consisted of either pneumothorax or hemothorax, while minor complications included asymptomatic stable or resolving pneumothoraces, transient hemoptysis or small hemothoraces. Conclusions: A combined procedure of percutaneous pulmonary core biopsy and stereotactic seed placement can be performed without additional risk of a major complication when compared to performing these separately.

  15. Core and shielding analysis of the SCM-100

    International Nuclear Information System (INIS)

    Olson, A. P.

    2002-01-01

    It is widely accepted that an intense neutron source can be produced in a suitable target by spallation neutrons generated by a high-current high-energy proton beam. Typical beam energy for such an accelerator is 400 to 2000 MeV. A conventional critical reactor can readily be replaced by a ''sub-critical reactor'' driven by this source. A 5 MW proton beam at 600 MeV can drive a sub-critical reactor to 100 MWt. The accelerator and the associated plant support equipment at these design specifications are complex systems, but they are well within recent technology. The purpose of this study was to examine core design and shielding design issues for a 100 MWt sodium-cooled fast-spectrum Sub-Critical Multiplier (SCM-100) based on LMFBR technology, but driven by an intense neutron source created by spallation reactions. SCM-100 is a component of the Accelerator Driven Test Facility. In this report we provide an overview of the SCM-100 concept. Two designs were investigated: (1) a vertical entry for the beam on the axial centerline; and (2) an inclined entry design where the core is ''C'' shaped and the beam enters the side of the target at an angle of 32 degrees. A brief overview of relevant shielding design data from EBR-II is also provided. The key result of this report is that the inclined entry design cannot achieve design objectives for radial power peaking. Consequently it cannot achieve design objectives for peak neutron flux. Axial power peaking factors are controlled by the axial fuel height and the axial reflector properties. These dimensions and compositions are very similar in SCM-100 to those of EBR-II. EBR-II had an axial power peaking factor of 1.093, and a radial power peaking factor of about 1.46. The radial power peaking of SCM-100 with the inclined entry is too extreme at 2.15, and cannot be made acceptable by modifying the size and detailed shape of the ''C'' shaped core and reflector. The axial power peaking of SCM-100 is very close to that of EBR

  16. Benchmarking Data Analysis and Machine Learning Applications on the Intel KNL Many-Core Processor

    OpenAIRE

    Byun, Chansup; Kepner, Jeremy; Arcand, William; Bestor, David; Bergeron, Bill; Gadepally, Vijay; Houle, Michael; Hubbell, Matthew; Jones, Michael; Klein, Anna; Michaleas, Peter; Milechin, Lauren; Mullen, Julie; Prout, Andrew; Rosa, Antonio

    2017-01-01

    Knights Landing (KNL) is the code name for the second-generation Intel Xeon Phi product family. KNL has generated significant interest in the data analysis and machine learning communities because its new many-core architecture targets both of these workloads. The KNL many-core vector processor design enables it to exploit much higher levels of parallelism. At the Lincoln Laboratory Supercomputing Center (LLSC), the majority of users are running data analysis applications such as MATLAB and O...

  17. Stress analysis of two-dimensional C/C composite components for HTGR's core restraint techanism

    International Nuclear Information System (INIS)

    Satoshi Hanawa; Taiju Shibata; Jyunya Sumita; Masahiro Ishihara; Tatsuo Iyoku; Kazuhiro Sawa

    2005-01-01

    Carbon fiber reinforced carbon matrix composite (C/C composite) is one of the most promising materials for HTGRs core components due to their high strength as well as high temperature resistibility. One of the most attractive applications of C/C composite is the core restraint mechanism. The core restraint mechanism is located around the reflector block and it works to tighten reactor core blocks so as to restrict un-supposition flow pass of coolant gas (bypass flow) in the core. The restriction of bypass flow reads to the high efficiency of coolant flow rate inside of the reactor core. For the future HTGRs and VHTR (Very High Temperature Reactor), it is important to develop the core restraint mechanism with C/C composite substitute for metallic materials as used for HTTR. For the application of C/C composite to core restraint mechanism, it is important to investigate the applicability of C/C composite in viewpoint of structural integrity. In the present study, supposing the application of 2D-C/C composite to core restraint mechanism, thermal stress behavior was analyzed by considering the thickness of the C/C composite and the gap between reflector block and core restraint. It was shown from the thermal stress analysis that the circumferential stress decreases with increasing the gap and that the restraint force increases with increasing the thickness. By optimizing the thickness of C/C composite and gap between reflector block and core restraint, the C/C composite is applicable to the core restraint mechanism. (authors)

  18. Error Analysis of High Frequency Core Loss Measurement for Low-Permeability Low-Loss Magnetic Cores

    DEFF Research Database (Denmark)

    Niroumand, Farideh Javidi; Nymand, Morten

    2016-01-01

    in magnetic cores is B-H loop measurement where two windings are placed on the core under test. However, this method is highly vulnerable to phase shift error, especially for low-permeability, low-loss cores. Due to soft saturation and very low core loss, low-permeability low-loss magnetic cores are favorable...... in many of the high-efficiency high power-density power converters. Magnetic powder cores, among the low-permeability low-loss cores, are very attractive since they possess lower magnetic losses in compared to gapped ferrites. This paper presents an analytical study of the phase shift error in the core...... loss measuring of low-permeability, low-loss magnetic cores. Furthermore, the susceptibility of this measurement approach has been analytically investigated under different excitations. It has been shown that this method, under square-wave excitation, is more accurate compared to sinusoidal excitation...

  19. Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

    International Nuclear Information System (INIS)

    Wang, S.J.; Chien, C.S.

    1996-01-01

    The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

  20. Geochemical analysis of core from a geothermal anomaly

    International Nuclear Information System (INIS)

    Haverslew, B.; Tammemagi, H.Y.

    1985-04-01

    A mild geothermal area in western Montana, USA, has been studied, as a natural analog, to learn about the effects that long-term heat generated by a repository containing spent nuclear fuel might have on the surrounding rock mass. The results of previous geological, geophysical and hydrogeological studies are briefly summarized. Extensive petrological studies have been undertaken on core samples obtained from a 2 km deep borehole drilled into the Empire Creek Stock. These include a detailed petrographic study, x-ray diffraction analyses, scanning electron microscope and electron microprobe analyses, porosity and permeability measurements, oxygen isotope analyses, uranium disequilibrium analyses and K-Ar age determinations. The implications to deep burial of nuclear wastes are discussed. 40 refs

  1. Thermal hydraulic analysis of the FFTF core using SUPERENERGY-2

    International Nuclear Information System (INIS)

    Cramer, E.R.; Basehore, K.L.

    1980-01-01

    SUPERENERGY-2 is the latest steady-state code in the ENERGY series, combining all of the desirable features of the previous ENERGY-I and SUPERENERGY versions in an optimized form. The result is an easily redimensionable, multiassembly code with many user-convenience features, such as automatic noding and a default constitutive package, that help minimize the effort and time associated with setting up large forced-convection problems. Improvements in physical modeling include generalized facial boundary conditions, duct wall gamma heating, and a model for double-ducted assemblies. The latter is used for modeling both multiduct test and absorber assemblies. SUPERENERGY-2 was used to calculate the temperature distribution in the first six rows of the FFTF core

  2. Biosynthesis of human sialophorins and analysis of the polypeptide core

    International Nuclear Information System (INIS)

    Remold-O'Donnell, E.; Kenney, D.; Rosen, F.S.

    1987-01-01

    Biosynthesis was examined of sialophorin (formerly called gpL115) which is altered in the inherited immunodeficiency Wiskott-Aldrich syndrome. Sialophorin is greater than 50% carbohydrate, primarily O-linked units of sialic acid, galactose, and galactosamine. Pulse-labeling with [ 35 S]methionine and chase incubation established that sialophorin is synthesized in CEM lymphoblastoid cells as an Mr 62,000 precursor which is converted within 45 min to mature glycosylated sialophorin, a long-lived molecule. Experiments with tunicamycin and endoglycosidase H demonstrated that sialophorin contains N-linked carbohydrate (approximately two units per molecule) and is therefore an N,O-glycoprotein. Pulse-labeling of tunicamycin-treated CEM cells together with immunoprecipitation provided the means to isolate the [ 35 S]-methionine-labeled polypeptide core of sialophorin and determine its molecular weight (58,000). This datum allowed us to express the previously established composition on a per molecule basis and determine that sialophorin molecules contain approximately 520 amino acid residues and greater than or equal to 100 O-linked carbohydrate units. A recent study showed that various blood cells express sialophorin and that there are two molecular forms: lymphocyte/monocyte sialophorin and platelet/neutrophil sialophorin. Biosynthesis of the two forms was compared by using sialophorin of CEM cells and sialophorin of MOLT-4 cells (another lymphoblastoid line) as models for lymphocyte/monocyte sialophorin and platelet/neutrophil sialophorin, respectively. The time course of biosynthesis and the content of N units were found to be identical for the two sialophorin species. [ 35 S]Methionine-labeled polypeptide cores of CEM sialophorin and MOLT sialophorin were isolated and compared by electrophoresis, isoelectrofocusing, and a newly developed peptide mapping technique

  3. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

    2002-01-01

    Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

  4. Development of three dimensional transient analysis code STTA for SCWR core

    International Nuclear Information System (INIS)

    Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

    2015-01-01

    Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

  5. SCDAP: a light water reactor computer code for severe core damage analysis

    International Nuclear Information System (INIS)

    Marino, G.P.; Allison, C.M.; Majumdar, D.

    1982-01-01

    Development of the first code version (MODO) of the Severe Core Damage Analysis Package (SCDAP) computer code is described, and calculations made with SCDAP/MODO are presented. The objective of this computer code development program is to develop a capability for analyzing severe disruption of a light water reactor core, including fuel and cladding liquefaction, flow, and freezing; fission product release; hydrogen generation; quenched-induced fragmentation; coolability of the resulting geometry; and ultimately vessel failure due to vessel-melt interaction. SCDAP will be used to identify the phenomena which control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and evaluation of severe fuel damage experiments and data. SCDAP/MODO addresses the behavior of a single fuel bundle. Future versions will be developed with capabilities for core-wide and vessel-melt interaction analysis

  6. Reconstruction and analysis of temperature and density spatial profiles inertial confinement fusion implosion cores

    International Nuclear Information System (INIS)

    Mancini, R. C.

    2007-01-01

    We discuss several methods for the extraction of temperature and density spatial profiles in inertial confinement fusion implosion cores based on the analysis of the x-ray emission from spectroscopic tracers added to the deuterium fuel. The ideas rely on (1) detailed spectral models that take into account collisional-radiative atomic kinetics, Stark broadened line shapes, and radiation transport calculations, (2) the availability of narrow-band, gated pinhole and slit x-ray images, and space-resolved line spectra of the core, and (3) several data analysis and reconstruction methods that include a multi-objective search and optimization technique based on a novel application of Pareto genetic algorithms to plasma spectroscopy. The spectroscopic analysis yields the spatial profiles of temperature and density in the core at the collapse of the implosion, and also the extent of shell material mixing into the core. Results are illustrated with data recorded in implosion experiments driven by the OMEGA and Z facilities

  7. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  8. Estimation of a Reactor Core Power Peaking Factor Using Support Vector Regression and Uncertainty Analysis

    International Nuclear Information System (INIS)

    Bae, In Ho; Naa, Man Gyun; Lee, Yoon Joon; Park, Goon Cherl

    2009-01-01

    The monitoring of detailed 3-dimensional (3D) reactor core power distribution is a prerequisite in the operation of nuclear power reactors to ensure that various safety limits imposed on the LPD and DNBR, are not violated during nuclear power reactor operation. The LPD and DNBR should be calculated in order to perform the two major functions of the core protection calculator system (CPCS) and the core operation limit supervisory system (COLSS). The LPD at the hottest part of a hot fuel rod, which is related to the power peaking factor (PPF, F q ), is more important than the LPD at any other position in a reactor core. The LPD needs to be estimated accurately to prevent nuclear fuel rods from melting. In this study, support vector regression (SVR) and uncertainty analysis have been applied to estimation of reactor core power peaking factor

  9. Analysis of impurity effect on Silicide fuels of the RSG-GAS core

    International Nuclear Information System (INIS)

    Tukiran-Surbakti

    2003-01-01

    Simulation of impurity effect on silicide fuel of the RSG-GAS core has been done. The aim of this research is to know impurity effect of the U-234 and U-236 isotopes in the silicide fuels on the core criticality. The silicide fuels of 250 g U loading and 19.75 of enrichment is used in this simulation. Cross section constant of fuels and non-structure material of core are generated by WIMSD/4 computer code, meanwhile impurity concentration was arranged from 0.01% to 2%. From the result of analysis can be concluded that the isotopes impurity in the fuels could make trouble in the core and the core can not be operated at critical after a half of its cycle length (350 MW D)

  10. Two dimensional dynamic analysis of sandwich plates with gradient foam cores

    Energy Technology Data Exchange (ETDEWEB)

    Mu, Lin; Xiao, Deng Bao; Zhao, Guiping [State Key Laboratory for Mechanical structure Strength and Vibration, School of AerospaceXi' an Jiaotong University, Xi' an (China); Cho, Chong Du [Dept. of Mechanical Engineering, Inha University, Inchon (Korea, Republic of)

    2016-09-15

    Present investigation is concerned about dynamic response of composite sandwich plates with the functionally gradient foam cores under time-dependent impulse. The analysis is based on a model of the gradient sandwich plate, in which the face sheets and the core adopt the Kirchhoff theory and a [2, 1]-order theory, respectively. The material properties of the gradient foam core vary continuously along the thickness direction. The gradient plate model is validated with the finite element code ABAQUS®. And the results show that the proposed model can predict well the free vibration of composite sandwich plates with gradient foam cores. The influences of gradient foam cores on the natural frequency, deflection and energy absorbing of the sandwich plates are also investigated.

  11. Study and analysis for the flow-induced vibration of the core barrel of a PWR

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan

    1989-01-01

    The resemblance criteria are derived and a test model is designed by applying the flow-soild coupling theory. After having completed the model analysis of the pressurized water reactor (PWR) core barrel in an 1:10 model, the dynamic characteristics are obtained. In an 1:5 reactor model with a hydraulic closed loop, the hydraulic vibration tests of the core barrel are performed, and the relations between the flow rate and the flow-induced pulse pressure on core barrel, acceleration and strain signals have been measured. The corresponding responses and a group of computational equations for hydraulic vibration are derived from these two experiments. The computational hydraulic vibration responses for core barrel in Qinshan Nuclear Power Plant are in good agreement with the test results, and it shows that the core barrel is safe within its lifetime of 30 years

  12. Manual of Standard Operating Procedures for Veterinary Drug Residue Analysis

    International Nuclear Information System (INIS)

    2016-01-01

    Laboratories are crucial to national veterinary drug residue monitoring programmes. However, one of the main challenges laboratories encounter is obtaining access to relevant methods of analysis. Thus, in addition to training, providing technical advice and transferring technology, the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture has resolved to develop clear and practical manuals to support Member State laboratories. The Coordinated Research Project (CRP) on Development of Radiometric and Allied Analytical Methods to Strengthen Residue Control Programs for Antibiotic and Anthelmintic Veterinary Drug Residues has developed a number of analytical methods as standard operating procedures (SOPs), which are now compiled here. This publication contains SOPs on chromatographic and spectrometric techniques, as well as radioimmunoassay and associated screening techniques, for various anthelmintic and antimicrobial veterinary drug residue analysis. Some analytical method validation protocols are also included. The publication is primarily aimed at food and environmental safety laboratories involved in testing veterinary drug residues, including under organized national residue monitoring programmes. It is expected to enhance laboratory capacity building and competence through the use of radiometric and complementary tools and techniques. The publication is also relevant for applied research on residues of veterinary drugs in food and environmental samples

  13. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  14. Automatization of an inverse surface temperature modelling procedure for Greenland ice cores, developed and evaluated using nitrogen and argon isotope data measured on the Gisp2 ice core

    Science.gov (United States)

    Döring, Michael; Kobashi, Takuro; Leuenberger, Markus

    2017-04-01

    In order to study Northern Hemisphere climate interactions and variability during the Holocene, access to high resolution surface temperature records of the Greenland ice sheet is an integral condition. Surface temperature reconstruction relies on firn densification combined with gas and heat diffusion [Severinghaus et al. (1998)]. In this study we use the model developed by Schwander et al. (1997). A theoretical δ15N record is generated for different temperature scenarios and compared with measurements by minimizing the mean squared error (MSE). The goal of the presented study is an automatization of this inverse modelling procedure. To solve the inverse problem, the Holocene temperature reconstruction is implemented in three steps. First a rough first guess temperature input (prior) is constructed which serves as the starting point for the optimization. Second, a smooth solution which transects the δ15N measurement data is generated following a Monte Carlo approach. It is assumed that the smooth solution contains all long term temperature trends and (together with the accumulation rate input) drives changes in firn column height, which generate the gravitational background signal in δ15N. Finally, the smooth solution is superimposed with high frequency information directly extracted from the δ15N measurement data. Following the approach, a high resolution Holocene temperature history for the Gisp2 site was extracted (posteriori), which leads to modelled δ15N data that fits the measurements in the low permeg level (MSE) and shows excellent agreement in timing and strength of the measurement variability. To evaluate the reconstruction procedure different synthetic data experiments were conducted underlining the quality of the method. Additionally, a second firn model [Goujon et al. (2003)] was used, which leads to very similar results, that shows the robustness of the presented approach. References: Goujon, C., Barnola, J.-M., Ritz, C. (2003). Modeling the

  15. HYDRATE CORE DRILLING TESTS

    Energy Technology Data Exchange (ETDEWEB)

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

  16. Extension of the analytic nodal diffusion solver ANDES to triangular-Z geometry and coupling with COBRA-IIIc for hexagonal core analysis

    International Nuclear Information System (INIS)

    Lozano, Juan-Andres; Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria

    2010-01-01

    In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal-hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthogonal to the triangle interfaces. The triangular nodalization avoids the singularities, that appear when applying transverse integration to hexagonal nodes, and allows the advantage of the mesh subdivision capabilities implicit within that geometry. As for the thermal-hydraulics, the extension of the coupling scheme to hexagonal geometry has been performed with the capability to model the core using either assembly-wise channels (hexagonal mesh) or a higher refinement with six channels per fuel assembly (triangular mesh). Achieving this level of TH mesh refinement with COBRA-IIIc code provides a better estimation of the in-core 3D flow distribution, improving the TH core modelling. The neutronics and thermal-hydraulics coupled code, ANDES/COBRA-IIIc, previously verified in Cartesian geometry core analysis, can also be applied now to full three-dimensional VVER core problems, as well as to other thermal and fast hexagonal core designs. Verification results are provided, corresponding to the different cases of the OECD/NEA-NSC VVER-1000 Coolant Transient Benchmarks.

  17. Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

    2017-01-01

    Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k_e_f_f and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

  18. Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Chenghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Shen, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2017-04-15

    Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k{sub eff} and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

  19. Toroidal HTS transformer with cold magnetic core - analysis with FEM software

    International Nuclear Information System (INIS)

    Grzesik, B; Stepien, M; Jez, R

    2010-01-01

    The aim of this paper is to present a thorough characterization of the toroidal HTS transformer by means of FEM analysis. The analysis was a 2D/3D harmonic electromagnetic and thermal analysis. The toroidal transformer operated in LN2 by being immersed together with the magnetic core in it, for which its power losses were acceptable. Two extreme variants of windings were analysed. The first one called parallel and the second called perpendicular. Three variants of the magnetic core were considered. In the first one the core was put outside of the windings, in the second the core was inside of the windings and in the third variant the core was outside as well as inside of the windings. The windings were made of HTS tape BiSCCO-2223/Ag while the magnetic core was made of the nanocrystalline material Finemet. The two windings, with a 1:1 turn-to-turn ratio, were uniformly distributed along the whole torus circumference. The output power, efficiency and power density are in the results of the analysis. The temperature distribution was also calculated. In summary, the performance of the transformer is better than those currently known.

  20. An analysis of cobalt irradiation in CANDU 6 reactor core

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Dumitrache, I.

    2003-01-01

    In CANDU reactors, one has the ability to replace the stainless steel adjuster rods with neutronically equivalent Co assemblies with a minimum impact on the power plant safety and efficiency. The 60 Co produced by 59 Co irradiation is used extensively in medicine and industry. The paper mainly describes some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronically equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and the heating of the irradiated cobalt rods are performed using the Monte Carlo codes MCNP5 and MONTEBURNS2.1. The 60 Co activity and heating evaluations are closely related to the neutronics computations and to the density evolution of cobalt isotopes during assumed in-core irradiation period. Unfortunately, the activities of these isotopes could not be evaluated directly using the burn-up capabilities of the MONTEBURNS code because of the lack of their neutron cross-section from the MCNP5 code library. Additional MCNP5 runs for all the cobalt assemblies have been done in order to compute the flux-spectrum, the 59 Co and the 60 Co radiative capture reaction rates in the adjusters. The 60m Co cross-section was estimated using the flux-spectrum and the ORIGEN2.1 code capabilities THERM and RES. These computational steps allowed the evaluation of the one-group cross-section for the radiative capture reactions of cobalt isotopes. The values obtained replaced the corresponding ones from the ORIGEN library, which have been estimated using the flux-spectrum specific to the fuel. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

  1. A SAS2H/KENO-V Methodology for 3D Full Core depletion analysis

    International Nuclear Information System (INIS)

    Milosevic, M.; Greenspan, E.; Vujic, J.; Petrovic, B.

    2003-04-01

    This paper describes the use of a SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrates its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. This approach reduces by more than an order of magnitude the time required for getting comparable results using the MOCUP code system. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark are in good agreement with the results of the ALPHA/PHOENIX/ANC code system. (author)

  2. Analysis of space-time core dynamics on reactor accident at Chernobyl

    International Nuclear Information System (INIS)

    Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

    1987-05-01

    Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

  3. Neutronics analysis of the TRIGA Mark II reactor core and its experimental facilities

    International Nuclear Information System (INIS)

    Khan, R.

    2010-01-01

    The neutronics analysis of the current core of the TRIGA Mark II research reactor is performed at the Atominstitute (ATI) of Vienna University of Technology. The current core is a completely mixed core having three different types of fuels i.e. aluminium clad 20 % enriched, stainless steel clad 20 % enriched and SS clad 70 % enriched (FLIP) Fuel Elements (FE(s)). The completely mixed nature and complicated irradiation history of the core makes the reactor physics calculations challenging. This PhD neutronics research is performed by employing the combination of two best and well practiced reactor simulation tools i.e. MCNP (general Monte Carlo N-particle transport code) for static analysis and ORIGEN2 (Oak Ridge Isotop Generation and depletion code) for dynamic analysis of the reactor core. The PhD work is started to develop a MCNP model of the first core configuration (March 1962) employing fresh fuel composition. The neutrons reaction data libraries ENDF/B-VI is applied taking the missing isotope of Samarium from JEFF3.1. The MCNP model of the very first core has been confirmed by three different local experiments performed on the first core configuration. These experiments include the first criticality, reactivity distribution and the neutron flux density distribution experiment. The first criticality experiment verifies the MCNP model that core achieves its criticality on addition of the 57th FE with a reactivity difference of about 9.3 cents. The measured reactivity worths of four FE(s) and a graphite element are taken from the log book and compared with MCNP simulated results. The percent difference between calculations and measurements ranges from 4 to 22 %. The neutron flux density mapping experiment confirms the model completely exhibiting good agreement between simulated and the experimental results. Since its first criticality, some additional 104-type and 110-type (FLIP) FE(s) have been added to keep the reactor into operation. This turns the current

  4. Analysis of excess reactivity of JOYO MK-III performance test core

    International Nuclear Information System (INIS)

    Maeda, Shigetaka; Yokoyama, Kenji

    2003-10-01

    JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

  5. Neutronic Analysis and Radiological Safety of RSG-GAS Reactor on 300 Grams Uranium Silicide Core

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Lily Suparlina; Rokhmadi

    2007-01-01

    As starting of usage silicide U 250 g fuel element in the core of RSG-GAS and will be continued with usage of silicide U 300 g fuel element, hence done beforehand neutronic analyse and radiological safety of RSG-GAS. Calculation done by ORIGEN2.1 code to calculate source term, and also by PC-COSYMA code to calculate radiological safety of radioactive dispersion from RSG-GAS. Calculation of radioactive dispersion done at condition of reactor is postulated be happened an accident of LOCA causing one fuel element to melt. Neutronic analysis indicate that silicide U 250 g full core shall to be operated beforehand during 625 MWD before converted to silicide U 300 g core. During operation of transition core with mixture of silicide U 250 g and 300 g, all parameter fulfill criterion of safety Designed Balance core of silicide U 300 g will be reached at the time of fifth full core. Result of calculation indicate that through mixture core of silicide U 250 and 300 g proposed can form silicide U 300 g balance core of reactor RSG-GAS safely. Calculation of radiology safety by deterministic for silicide U 300 g balance core, and accident postulation which is equal to core of silicide U 250 g yield output in the form of radiation activity (radionuclide concentration in the air and deposition on the ground), radiation dose (collective and individual), radiation effect (short- and long-range), which accepted by society in each perceived sector. Result of calculation indicated that dose accepted by society is not pass permitted boundary for public society if happened accident. (author)

  6. Development of flow network analysis code for block type VHTR core by linear theory method

    International Nuclear Information System (INIS)

    Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

    2012-01-01

    VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

  7. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  8. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    Battum, E. van; Serov, I.V.

    1993-07-01

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  9. A Benchmark Study of a Seismic Analysis Program for a Single Column of a HTGR Core

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A seismic analysis program, SAPCOR (Seismic Analysis of Prismatic HTGR Core), was developed in Korea Atomic Energy Research Institute. The program is used for the evaluation of deformed shapes and forces on the graphite blocks which using point-mass rigid bodies with Kelvin-Voigt impact models. In the previous studies, the program was verified using theoretical solutions and benchmark problems. To validate the program for more complicated problems, a free vibration analysis of a single column of a HTGR core was selected and the calculation results of the SAPCOR and a commercial FEM code, Abaqus, were compared in this study.

  10. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    Ravnik, M.

    1988-11-01

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  11. TMI-2 core debris grab samples: Examination and analysis: Part 1

    International Nuclear Information System (INIS)

    Akers, D.W.; Carlson, E.R.; Cook, B.A.; Ploger, S.A.; Carlson, J.O.

    1986-09-01

    Six samples of particulate debris were removed from the TMI-2 core rubble bed during September and October 1983, and five more samples were obtained in March 1984. The samples (up to 174 g each) were obtained at two locations in the core: H8 (center) and E9 (mid-radius). Ten of the eleven samples were examined at the Idaho National Engineering Laboratory to obtain data on the physical and chemical nature of the debris and the postaccident condition of the core. Portions of the samples also were subjected to differential thermal analysis at Rockwell Hanford Operations and metallurgical and chemical examinations at Argonne National Laboratories. This report presents results of the examination of the core debris grab samples, including physical, metallurgical, chemical, and radiochemical analyses. The results indicate that temperatures in the core reached at least 3100 K during the TMI-2 accident, fuel melting and significant mixing of core structural material occurred, and large fractions of some radionuclides (e.g., 90 Sr and 144 Ce) were retained in the core

  12. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  13. Radiation and environmental data analysis computer (REDAC) hardware, software band analysis procedures

    International Nuclear Information System (INIS)

    Hendricks, T.J.

    1985-01-01

    The REDAC was conceived originally as a tape verifier for the Radiation and Environmental Data Acquisition Recorder (REDAR). From that simple beginning in 1971, the REDAC has evolved into a family of systems used for complete analysis of data obtained by the REDAR and other acquisition systems. Portable or mobile REDACs are deployed to support checkout and analysis tasks in the field. Laboratory systems are additionally used for software development, physics investigations, data base management and graphics. System configurations range from man-portable systems to a large laboratory-based system which supports time-shared analysis and development tasks. Custom operating software allows the analyst to process data either interactively or by batch procedures. Analysis packages are provided for numerous necessary functions. All these analysis procedures can be performed even on the smallest man-portable REDAC. Examples of the multi-isotope stripping and radiation isopleth mapping are presented. Techniques utilized for these operations are also presented

  14. A finite volume procedure for fluid flow, heat transfer and solid-body stress analysis

    KAUST Repository

    Jagad, P. I.; Puranik, B. P.; Date, A. W.

    2018-01-01

    A unified cell-centered unstructured mesh finite volume procedure is presented for fluid flow, heat transfer and solid-body stress analysis. An in-house procedure (A. W. Date, Solution of Transport Equations on Unstructured Meshes with Cell

  15. Quantitative analysis of core fucosylation of serum proteins in liver diseases by LC-MS-MRM.

    Science.gov (United States)

    Ma, Junfeng; Sanda, Miloslav; Wei, Renhuizi; Zhang, Lihua; Goldman, Radoslav

    2018-02-07

    Aberrant core fucosylation of proteins has been linked to liver diseases. In this study, we carried out multiple reaction monitoring (MRM) quantification of core fucosylated N-glycopeptides of serum proteins partially deglycosylated by a combination of endoglycosidases (endoF1, endoF2, and endoF3). To minimize variability associated with the preparatory steps, the analysis was performed without enrichment of glycopeptides or fractionation of serum besides the nanoRP chromatography. Specifically, we quantified core fucosylation of 22 N-glycopeptides derived from 17 proteins together with protein abundance of these glycoproteins in a cohort of 45 participants (15 disease-free control, 15 fibrosis and 15 cirrhosis patients) using a multiplex nanoUPLC-MS-MRM workflow. We find increased core fucosylation of 5 glycopeptides at the stage of liver fibrosis (i.e., N630 of serotransferrin, N107 of alpha-1-antitrypsin, N253 of plasma protease C1 inhibitor, N397 of ceruloplasmin, and N86 of vitronectin), increase of additional 6 glycopeptides at the stage of cirrhosis (i.e., N138 and N762 of ceruloplasmin, N354 of clusterin, N187 of hemopexin, N71 of immunoglobulin J chain, and N127 of lumican), while the degree of core fucosylation of 10 glycopeptides did not change. Interestingly, although we observe an increase in the core fucosylation at N86 of vitronectin in liver fibrosis, core fucosylation decreases on the N169 glycopeptide of the same protein. Our results demonstrate that the changes in core fucosylation are protein and site specific during the progression of fibrotic liver disease and independent of the changes in the quantity of N-glycoproteins. It is expected that the fully optimized multiplex LC-MS-MRM assay of core fucosylated glycopeptides will be useful for the serologic assessment of the fibrosis of liver. We have quantified the difference in core fucosylation among three comparison groups (healthy control, fibrosis and cirrhosis patients) using a sensitive and

  16. Procedural-support music therapy in the healthcare setting: a cost-effectiveness analysis.

    Science.gov (United States)

    DeLoach Walworth, Darcy

    2005-08-01

    This comparative analysis examined the cost-effectiveness of music therapy as a procedural support in the pediatric healthcare setting. Many healthcare organizations are actively attempting to reduce the amount of sedation for pediatric patients undergoing various procedures. Patients receiving music therapy-assisted computerized tomography scans ( n = 57), echocardiograms ( n = 92), and other procedures ( n = 17) were included in the analysis. Results of music therapy-assisted procedures indicate successful elimination of patient sedation, reduction in procedural times, and decrease in the number of staff members present for procedures. Implications for nurses and music therapists in the healthcare setting are discussed.

  17. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  18. Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

    International Nuclear Information System (INIS)

    Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

    1988-07-01

    Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

  19. Prompt Gamma Activation Analysis of the Nyírlugos obsidian core depot find

    Directory of Open Access Journals (Sweden)

    Zsolt Kasztovszky

    2014-03-01

    Full Text Available The Nyírlugos obsidian core depot find is one of the most important lithic assemblages in the collection of the Hungarian National Museum (HNM. The original set comprised 12 giant obsidian cores, of which 11 are currently on the permanent archaeological exhibition of the HNM. One of the cores is known to be inDebrecen. The first publication attributed the hoard, on the strength of giant (flint blades known from the Early and Middle Copper Age Tiszapolgár and Bodrogkeresztúr cultures, to the Copper Age. In the light of recent finds it is more likely to belong to the Middle Neolithic period. The source area was defined as Tokaj Mts., about100 kmto the NW from Nyírlugos. The size and beauty of the exceptional pieces exclude any invasive analysis. Using Prompt Gamma Activation Analysis (PGAA, we can measure major chemical components and some key trace elements of stone artefacts with adequate accuracy to successfully determine provenance of obsidian. Recent methodological development also facilitated the study of relatively large objects like the Nyírlugos cores. The cores were individually measured by PGAA. The results show that the cores originate from the Carpathian 1 sources, most probably the Viničky variety (C1b. The study of the hoard as a batch is an important contribution to the assessment of prehistoric trade and allows us to reconsider the so-called Carpathian, especially Carpathian 1 (Slovakian sources.

  20. Core psychopathology in anorexia nervosa and bulimia nervosa: A network analysis.

    Science.gov (United States)

    Forrest, Lauren N; Jones, Payton J; Ortiz, Shelby N; Smith, April R

    2018-04-25

    The cognitive-behavioral theory of eating disorders (EDs) proposes that shape and weight overvaluation are the core ED psychopathology. Core symptoms can be statistically identified using network analysis. Existing ED network studies support that shape and weight overvaluation are the core ED psychopathology, yet no studies have estimated AN core psychopathology and concerns exist about the replicability of network analysis findings. The current study estimated ED symptom networks among people with anorexia nervosa (AN) and bulimia nervosa (BN) and among a combined group of people with AN and BN. Participants were girls and women with AN (n = 604) and BN (n = 477) seeking residential ED treatment. ED symptoms were assessed with the Eating Disorder Examination-Questionnaire (EDE-Q); 27 of the EDE-Q items were included as nodes in symptom networks. Core symptoms were determined by expected influence and strength values. In all networks, desiring weight loss, restraint, shape and weight preoccupation, and shape overvaluation emerged as the most important symptoms. In addition, in the AN and combined networks, fearing weight gain emerged as an important symptom. In the BN network, weight overvaluation emerged as another important symptom. Findings support the cognitive-behavioral premise that shape and weight overvaluation are at the core of AN psychopathology. Our BN and combined network findings provide a high degree of replication of previous findings. Clinically, findings highlight the importance of considering shape and weight overvaluation as a severity specifier and primary treatment target for people with EDs. © 2018 Wiley Periodicals, Inc.

  1. A method for statistical steady state thermal analysis of reactor cores

    International Nuclear Information System (INIS)

    Whetton, P.A.

    1980-01-01

    This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes. Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved. (orig.)

  2. Study and analysis on the flow induced vibration of the core barrel of PWR

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan; Peng YongYong; Zhang Huijun; Wang Yufen; Xie Yongcheng; Guo Chunhua; Shen Qinping

    1989-01-01

    The deduction of the resemblance criterion and the design of the test model by applying flow-solid coupling theory are described. The model analysis of a core barrel both in the air and stationary water were performed in a 1:10 model, thus obtaining the dynamic characteristic. In a 1:5 reactor model with a hydraulic closed loop, the inner structure and support were modeled for performing hydraulic closed loop, the inner structure and support were modeled for performing hydraulic vibration test of the core barrel. The flow induced pulse pressure of the core barrel and corresponding response were obtained by using miniature pressure capsule, strain gauge and accelerometer. Power spectrum, correlation functions, transfer function and amplitudes under different flow velocities were calculated. The hydraulic vibration test shows that the core barrel will be in safety during its 30-year life time

  3. Numerical analysis of temperature fluctuation in core outlet region of China experimental fast reactor

    International Nuclear Information System (INIS)

    Zhu Huanjun; Xu Yijun

    2014-01-01

    The temperature fluctuation in core outlet region of China Experimental Fast Reactor (CEFR) was numerically simulated by the CFD software Star CCM+. With the core outlet temperatures, flows etc. under rated conditions given as boundary conditions, a 1/4 region model of the reactor core outlet region was established and calculated using LES method for this problem. The analysis results show that while CEFR operates under rated conditions, the temperature fluctuation in lower part of core outlet region is mainly concentrated in area over the edge components (steel components, control rod assembly), and one in upper part is remarkable in area above all the components. The largest fluctuation amplitude is 19 K and the remarkable frequency is below 5 Hz, and it belongs to typically low frequency fluctuation. The conclusion is useful for further experimental work. (authors)

  4. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  5. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    Schubert, J.D.; Gulden, W.; Jacobs, G.; Meyder, R.; Sengpiel, W.

    1985-05-01

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP) [de

  6. Safety analysis of RSG-GAS Silicide core using one line cooling system

    International Nuclear Information System (INIS)

    Endiah-Puji-Hastuti

    2003-01-01

    In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor has been determined and to continuing this program, steady state and transient analysis were done. The analysis was done by means of a core thermal hydraulic code, COOLOD-N, and PARET. The codes solves core thermal hydraulic equation at steady state conditions and transient, respectively. By using silicide core data and coast down flow rate as the input, thermal hydraulics parameters such as fuel cladding and fuel meat temperatures as well as safety margin against flow instability were calculated. Imposing the safety criteria to the results of steady state and transient analysis, maximum permissible power for this operation was obtained as much as 17.1 MW

  7. A High-Resolution Continuous Flow Analysis System for Polar Ice Cores

    DEFF Research Database (Denmark)

    Dallmayr, Remi; Goto-Azuma, Kumiko; Kjær, Helle Astrid

    2016-01-01

    of Polar Research (NIPR) in Tokyo. The system allows the continuous analysis of stable water isotopes and electrical conductivity, as well as the collection of discrete samples from both inner and outer parts of the core. This CFA system was designed to have sufficiently high temporal resolution to detect...... signals of abrupt climate change in deep polar ice cores. To test its performance, we used the system to analyze different climate intervals in ice drilled at the NEEM (North Greenland Eemian Ice Drilling) site, Greenland. The quality of our continuous measurement of stable water isotopes has been......In recent decades, the development of continuous flow analysis (CFA) technology for ice core analysis has enabled greater sample throughput and greater depth resolution compared with the classic discrete sampling technique. We developed the first Japanese CFA system at the National Institute...

  8. Gap analysis: a method to assess core competency development in the curriculum.

    Science.gov (United States)

    Fater, Kerry H

    2013-01-01

    To determine the extent to which safety and quality improvement core competency development occurs in an undergraduate nursing program. Rapid change and increased complexity of health care environments demands that health care professionals are adequately prepared to provide high quality, safe care. A gap analysis compared the present state of competency development to a desirable (ideal) state. The core competencies, Nurse of the Future Nursing Core Competencies, reflect the ideal state and represent minimal expectations for entry into practice from pre-licensure programs. Findings from the gap analysis suggest significant strengths in numerous competency domains, deficiencies in two competency domains, and areas of redundancy in the curriculum. Gap analysis provides valuable data to direct curriculum revision. Opportunities for competency development were identified, and strategies were created jointly with the practice partner, thereby enhancing relevant knowledge, attitudes, and skills nurses need for clinical practice currently and in the future.

  9. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    International Nuclear Information System (INIS)

    Oguma, R.

    1998-04-01

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

  10. Probabilistic safety analysis procedures guide, Sections 8-12. Volume 2, Rev. 1

    International Nuclear Information System (INIS)

    McCann, M.; Reed, J.; Ruger, C.; Shiu, K.; Teichmann, T.; Unione, A.; Youngblood, R.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. The first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. This second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  11. Probabilistic safety analysis procedures guide. Sections 1-7 and appendices. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Bari, R.A.; Buslik, A.J.; Cho, N.Z.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. This first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. The second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  12. Head-camera video recordings of trauma core competency procedures can evaluate surgical resident's technical performance as well as colocated evaluators.

    Science.gov (United States)

    Mackenzie, Colin F; Pasley, Jason; Garofalo, Evan; Shackelford, Stacy; Chen, Hegang; Longinaker, Nyaradzo; Granite, Guinevere; Pugh, Kristy; Hagegeorge, George; Tisherman, Samuel A

    2017-07-01

    Unbiased evaluation of trauma core competency procedures is necessary to determine if residency and predeployment training courses are useful. We tested whether a previously validated individual procedure score (IPS) for individual procedure vascular exposure and fasciotomy (FAS) performance skills could discriminate training status by comparing IPS of evaluators colocated with surgeons to blind video evaluations. Performance of axillary artery (AA), brachial artery (BA), and femoral artery (FA) vascular exposures and lower extremity FAS on fresh cadavers by 40 PGY-2 to PGY-6 residents was video-recorded from head-mounted cameras. Two colocated trained evaluators assessed IPS before and after training. One surgeon in each pretraining tertile of IPS for each procedure was randomly identified for blind video review. The same 12 surgeons were video-recorded repeating the procedures less than 4 weeks after training. Five evaluators independently reviewed all 96 randomly arranged deidentified videos. Inter-rater reliability/consistency, intraclass correlation coefficients were compared by colocated versus video review of IPS, and errors. Study methodology and bias were judged by Medical Education Research Study Quality Instrument and the Quality Assessment of Diagnostic Accuracy Studies criteria. There were no differences (p ≥ 0.5) in IPS for AA, FA, FAS, whether evaluators were colocated or reviewed video recordings. Evaluator consistency was 0.29 (BA) - 0.77 (FA). Video and colocated evaluators were in total agreement (p = 1.0) for error recognition. Intraclass correlation coefficient was 0.73 to 0.92, dependent on procedure. Correlations video versus colocated evaluations were 0.5 to 0.9. Except for BA, blinded video evaluators discriminated (p competency. Prognostic study, level II.

  13. Analysis of decision procedures for a sequence of inventory periods

    International Nuclear Information System (INIS)

    Avenhaus, R.

    1982-07-01

    Optimal test procedures for a sequence of inventory periods will be discussed. Starting with a game theoretical description of the conflict situation between the plant operator and the inspector, the objectives of the inspector as well as the general decision theoretical problem will be formulated. In the first part the objective of 'secure' detection will be emphasized which means that only at the end of the reference time a decision is taken by the inspector. In the second part the objective of 'timely' detection will be emphasized which will lead to sequential test procedures. At the end of the paper all procedures will be summarized, and in view of the multitude of procedures available at the moment some comments about future work will be given. (orig./HP) [de

  14. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

  15. A method for statistical steady state thermal analysis of reactor cores

    International Nuclear Information System (INIS)

    Whetton, P.A.

    1981-01-01

    In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)

  16. Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

    International Nuclear Information System (INIS)

    Shemon, Emily R.

    2016-01-01

    Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact

  17. Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-10

    Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact

  18. Development of the Monju core safety analysis numerical models by super-COPD code

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Minami, Masaki

    2010-12-01

    Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model. The numerical model for core safety analysis has been built based on the best estimate model validated by the actually measured plant behavior up to 40% rated power conditions, taking over safety analysis models of conventionally applied COPD and HARHO-IN codes, to be capable of overall calculations of the entire plant with the safety protection and control systems. Among the constituents of the analytical model, neutronic-thermal model, heat transfer and hydraulic models of PHTS, SHTS, and water/steam system are individually verified by comparisons with the conventional calculations. Comparisons are also made with the actually measured plant behavior up to 40% rated power conditions to confirm the calculation adequacy and conservativeness of the input data. The unified analytical model was applied to analyses of in total 8 anomaly events; reactivity insertion, abnormal power distribution, decrease and increase of coolant flow rate in PHTS, SHTS and water/steam systems. The resulting maximum values and temporal variations of the key parameters in safety evaluation; temperatures of fuel, cladding, in core sodium coolant and RV inlet and outlet coolant have negligible discrepancies against the existing analysis result in the annex to the construction permit application, verifying the unified analytical model. These works have enabled analytical evaluation of Monju core heat removal capability by Super-COPD utilizing the

  19. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-05-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

  20. New non-cognitive procedures for medical applicant selection: a qualitative analysis in one school.

    Science.gov (United States)

    Katz, Sara; Vinker, Shlomo

    2014-11-07

    Recent data have called into question the reliability and predictive validity of standard admission procedures to medical schools. Eliciting non-cognitive attributes of medical school applicants using qualitative tools and methods has thus become a major challenge. 299 applicants aged 18-25 formed the research group. A set of six research tools was developed in addition to the two existing ones. These included: a portfolio task, an intuitive task, a cognitive task, a personal task, an open self-efficacy questionnaire and field-notes. The criteria-based methodology design used constant comparative analysis and grounded theory techniques to produce a personal attributes profile per participant, scored on a 5-point scale holistic rubric. Qualitative validity of data gathering was checked by comparing the profiles elicited from the existing interview against the profiles elicited from the other tools, and by comparing two profiles of each of the applicants who handed in two portfolio tasks. Qualitative validity of data analysis was checked by comparing researcher results with those of an external rater (n =10). Differences between aggregated profile groups were checked by the Npar Wilcoxon Signed Ranks Test and by Spearman Rank Order Correlation Test. All subjects gave written informed consent to their participation. Privacy was protected by using code numbers. A concept map of 12 personal attributes emerged, the core constructs of which were motivation, sociability and cognition. A personal profile was elicited. Inter-rater agreement was 83.3%. Differences between groups by aggregated profiles were found significant (p < .05, p < .01, p < .001).A random sample of sixth year students (n = 12) underwent the same admission procedure as the research group. Rank order was different; and arrogance was a new construct elicited in the sixth year group. This study suggests a broadening of the methodology for selecting medical school applicants. This methodology

  1. Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Nabbi, R.

    1987-02-01

    In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

  2. Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

    2007-07-01

    Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

  3. Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

    International Nuclear Information System (INIS)

    Tukiran

    2002-01-01

    The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

  4. Experimental and numerical analysis of fluid - structure interaction effects in a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.; Melloni, R.; Paoluzzi, R.; Bonacina, G.; Castoldi, A.; Zola, M.

    1990-01-01

    Dynamic experiments in air and water (simulating liquid sodium) were performed by ISMES, on behalf of ENEA, on various core element groups of the Italian PEC fast reactor. Bundles of one, seven and nineteen mock-ups reproducing fuel, reflecting and neutron shield elements in full scale were analysed on shaking tables. Tests concerned both groups of equal elements and mixed configurations which corresponded to real core parts. The effects of PEC core-restraint ring were also studied. Seismic excitations of up to 2.5 g were applied to core diagrid. Test results were analysed by use of the one-dimensional program CORALIE and the two-dimensional program CLASH. The study allowed the fluid effects in the PEC core to be evaluated; it also contributed to validation of the above mentioned programs for their general use for fast reactor core analysis. This paper presents the main features of the experimental and the numerical studies and reports comparisons between calculations and measurements. (author)

  5. Analysis of pan-genome to identify the core genes and essential genes of Brucella spp.

    Science.gov (United States)

    Yang, Xiaowen; Li, Yajie; Zang, Juan; Li, Yexia; Bie, Pengfei; Lu, Yanli; Wu, Qingmin

    2016-04-01

    Brucella spp. are facultative intracellular pathogens, that cause a contagious zoonotic disease, that can result in such outcomes as abortion or sterility in susceptible animal hosts and grave, debilitating illness in humans. For deciphering the survival mechanism of Brucella spp. in vivo, 42 Brucella complete genomes from NCBI were analyzed for the pan-genome and core genome by identification of their composition and function of Brucella genomes. The results showed that the total 132,143 protein-coding genes in these genomes were divided into 5369 clusters. Among these, 1710 clusters were associated with the core genome, 1182 clusters with strain-specific genes and 2477 clusters with dispensable genomes. COG analysis indicated that 44 % of the core genes were devoted to metabolism, which were mainly responsible for energy production and conversion (COG category C), and amino acid transport and metabolism (COG category E). Meanwhile, approximately 35 % of the core genes were in positive selection. In addition, 1252 potential essential genes were predicted in the core genome by comparison with a prokaryote database of essential genes. The results suggested that the core genes in Brucella genomes are relatively conservation, and the energy and amino acid metabolism play a more important role in the process of growth and reproduction in Brucella spp. This study might help us to better understand the mechanisms of Brucella persistent infection and provide some clues for further exploring the gene modules of the intracellular survival in Brucella spp.

  6. Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man

    2007-01-01

    This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

  7. Development of Uncertainty Analysis Method for SMART Digital Core Protection and Monitoring System

    International Nuclear Information System (INIS)

    Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

    2012-01-01

    The Korea Atomic Energy Research Institute has developed a system-integrated modular advanced reactor (SMART) for a seawater desalination and electricity generation. Online digital core protection and monitoring systems, called SCOPS and SCOMS respectively were developed. SCOPS calculates minimum DNBR and maximum LPD based on the several online measured system parameters. SCOMS calculates the variables of limiting conditions for operation. KAERI developed overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system. By applying overall uncertainty factors in on-line SCOPS/SCOMS calculation, calculated LPD and DNBR are conservative with a 95/95 probability/confidence level. In this paper, uncertainty analysis method is described for SMART core protection and monitoring system

  8. DNBR calculation in digital core protection system by a subchannel analysis code

    International Nuclear Information System (INIS)

    In, W. K.; Yoo, Y. J.; Hwang, T. H.; Ji, S. K.

    2001-01-01

    The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR

  9. Analysis of severe core damage accident progression for the heavy water reactor

    International Nuclear Information System (INIS)

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

  10. Development of seismic analysis model for HTGR core on commercial FEM code

    International Nuclear Information System (INIS)

    Tsuji, Nobumasa; Ohashi, Kazutaka

    2015-01-01

    The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

  11. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    International Nuclear Information System (INIS)

    Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

    1983-01-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  12. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    Energy Technology Data Exchange (ETDEWEB)

    Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  13. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  14. Analysis of subchannel effects and their treatment in average channel PWR core models

    International Nuclear Information System (INIS)

    Cuervo, D.; Ahnert, C.; Aragones, J.M.

    2004-01-01

    Neutronic thermal-hydraulic coupling is meanly made at this moment using whole plant thermal-hydraulic codes with one channel per assembly or quarter of assembly in more detailed cases. To extract safety limits variables a new calculation has to be performed using thermal-hydraulic subchannel codes in an embedded or off-line manner what implies an increase of calculation time. Another problem of this separated analysis of whole core and not channel is that the whole core calculation is not resolving the real problem due to the modification of the variables values by the homogenization process that is carried out to perform the whole core analysis. This process is making that some magnitudes are over or under-predicted causing that the problem that is being solved is not the original one. The purpose of the work that is being developed is to investigate the effects of the averaging process in the results obtained by the whole core analysis and to develop some corrections that may be included in this analysis to obtain results closer to the ones obtained by a detailed subchannel analysis. This paper shows the results obtained for a sample case and the conclusions for future work. (author)

  15. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  16. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    Setiyanto

    2014-01-01

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  17. Solution Tree Problem Solving Procedure for Engineering Analysis ...

    African Journals Online (AJOL)

    Illustrations are provided in the thermofluid engineering area to showcase the procedure's applications. This approach has proved to be a veritable tool for enhancing the problem-solving and computer algorithmic skills of engineering students, eliciting their curiosity, active participation and appreciation of the taught course.

  18. A Comparative Analysis of the Procedure Employed in Item ...

    African Journals Online (AJOL)

    Zimbabwe Journal of Educational Research ... and psychological scales designed to measure constructs in education and social sciences were purposively selected for the study based on accessibility and availability of validation information. The instruments used for the study were scaling procedures used in 27 published ...

  19. Analysis of emergency response procedures and air traffic accidents ...

    African Journals Online (AJOL)

    Incessant air transport accidents have been a source of concern to stakeholders and aviation experts in Nigeria, yet the response and process has not been adequately appraised. This study attempts an evaluation of the emergency response procedures in the aviation industry with particular focus on Murtala Muhammed ...

  20. Enhancements to the Image Analysis Tool for Core Punch Experiments and Simulations (vs. 2014)

    Energy Technology Data Exchange (ETDEWEB)

    Hogden, John Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-06

    A previous paper (Hogden & Unal, 2012, Image Analysis Tool for Core Punch Experiments and Simulations) described an image processing computer program developed at Los Alamos National Laboratory. This program has proven useful so developement has been continued. In this paper we describe enhacements to the program as of 2014.

  1. Microscopic analysis on showers recorded as single core on X-ray films

    International Nuclear Information System (INIS)

    Amato, N.M.; Arata, N.; Maldonado, R.H.C.

    1983-01-01

    Cosmic-ray particles recorded as single dark spots on X-ray films with use of the emulsion chamber data of Brazil-Japan Collaboration are studied. Some results of microscopic analysis of such single-core-like showers on nuclear emulsion plates are reported. (Author) [pt

  2. Direct chemical analysis of frozen ice cores by UV-laser ablation ICPMS

    DEFF Research Database (Denmark)

    Müller, Wolfgang; Shelley, J. Michael G.; Rasmussen, Sune Olander

    2011-01-01

    Cryo-cell UV-LA-ICPMS is a new technique for direct chemical analysis of frozen ice cores at high spatial resolution (dust records and annual layer signatures at unprecedented spatial/time resolution. Uniquely......, the location of cation impurities relative to grain boundaries in recrystallized ice can be assessed....

  3. Core ADHD Symptom Improvement with Atomoxetine versus Methylphenidate: A Direct Comparison Meta-Analysis

    Science.gov (United States)

    Hazell, Philip L.; Kohn, Michael R.; Dickson, Ruth; Walton, Richard J.; Granger, Renee E.; van Wyk, Gregory W.

    2011-01-01

    Objective: Previous studies comparing atomoxetine and methylphenidate to treat ADHD symptoms have been equivocal. This noninferiority meta-analysis compared core ADHD symptom response between atomoxetine and methylphenidate in children and adolescents. Method: Selection criteria included randomized, controlled design; duration 6 weeks; and…

  4. Overview of current RFSP-code capabilities for CANDU core analysis

    International Nuclear Information System (INIS)

    Rouben, B.

    1996-01-01

    RFSP (Reactor Fuelling Simulation Program) is the major finite-reactor computer code in use at the Atomic Energy of Canada Limited for the design and analysis of CANDU reactor cores. An overview is given of the major computational capabilities available in RFSP. (author) 11 refs., 29 figs

  5. On the Paleostress Analysis Using Kinematic Indicators Found on an Oriented Core

    Czech Academy of Sciences Publication Activity Database

    Nováková, Lucie; Brož, Milan

    2014-01-01

    Roč. 2, č. 2 (2014), s. 76-83 ISSN 2331-9593 R&D Projects: GA MPO(CZ) FR-TI1/367 Institutional support: RVO:67985891 Keywords : paleostress analysis * borehole core * kinematic indicators * bias sampling * recent stress Subject RIV: DC - Siesmology, Volcanology, Earth Structure http://www.hrpub.org/download/20140105/UJG6-13901884.pdf

  6. Methodology for reactor core physics analysis - part 2; Metodologia de analise fisica do nucleo - etapa 2

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni Filho, P; Fernandes, V B; Lima Bezerra, J de; Santos, T I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs.

  7. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  8. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

    2012-01-01

    Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000 ® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

  9. Analysis of assistance procedures to normal birth in primiparous

    Directory of Open Access Journals (Sweden)

    Joe Luiz Vieira Garcia Novo

    2016-04-01

    Full Text Available Introduction: Current medical technologies in care in birth increased maternal and fetal benefits persist, despite numerous unnecessary procedures. The purpose of the normal childbirth care is to have healthy women and newborns, using a minimum of safe interventions. Objective: To analyze the assistance to normal delivery in secondary care maternity. Methodology: A total of 100 primiparous mothers who had vaginal delivery were included, in which care practices used were categorized: 1 according to the WHO classification for assistance to normal childbirth: effective, harmful, used with caution and used inappropriately; 2 associating calculations with the Bologna Index parameters: presence of a birth partner, partograph, no stimulation of labor, delivery in non-supine position, and mother-newborn skin-to-skin contact. Results: Birth partners (85%, correctly filled partographs (62%, mother-newborn skin-to-skin contact (36%, use of oxytocin (87%, use of parenteral nutrition during labor (86% and at delivery (74%, episiotomy (94% and uterine fundal pressure in the expulsion stage (58%. The overall average value of the Bologna Index of the mothers analyzed was 1.95. Conclusions: Some effective procedures recommended by WHO (presence of a birth partner, some effective and mandatory practices were not complied with (partograph completely filled, potentially harmful or ineffective procedures were used (oxytocin in labor/post-partum, as well as inadequate procedures (uterine fundal pressure during the expulsion stage, use of forceps and episiotomy. The maternity’s care model did not offer excellence procedures in natural birth to their mothers in primiparity, (BI=1.95.

  10. EFIT fuel cycle analysis with the EQL3D procedure

    International Nuclear Information System (INIS)

    Krepel, Jiri; Mikityuk, Konstantin; Sarotto, Massimo; Artioli, Carlo

    2009-01-01

    Accelerator Driven Systems (ADS) represent one of the possible future strategies for Minor Actinides (MA) transmutation. EFIT - European Facility for Industrial Transmutation is a 400 MWth ADS designed in the future of EUROTRANS project. It is fuelled by MA and Pu embedded in the inert Mg matrix, cooled by lead (673-753 K), and driven by an accelerator, which provides 15 mA current of 800 MeV protons. The subcritical core is divided into three radial zones, which differ in pin diameter or inert matrix percentage. This design flattens the flux distribution and enables to maximize the power density. (author)

  11. Formation evaluation in Devonian shale through application of new core and log analysis methods

    International Nuclear Information System (INIS)

    Luffel, D.L.; Guidry, F.K.

    1990-01-01

    In the Devonian shale of the Appalachian Basin all porosity in excess of about 2.5 percent is generally occupied by free hydrocarbons, which is mostly gas, based on results of new core and log analysis methods. In this study, sponsored by the Gas Research Institute, reservoir porosities averaged about 5 percent and free gas content averaged about 2 percent by bulk volume, based on analyses on 519 feet of conventional core in four wells. In this source-rich Devonian shale, which also provides the reservoir storage, the rock everywhere appears to be at connate, or irreducible, water saturation corresponding to two or three percent of bulk volume. This became evident when applying the new core and log analysis methods, along with a new plotting method relating bulk volume of pore fluids to porosity. This plotting method has proved to be a valuable tool: it provides useful insight on the fluid distribution present in the reservoir, it provides a clear idea of porosity required to store free hydrocarbons, it leads to a method of linking formation factor to porosity, and it provides a good quality control method to monitor core and log analysis results. In the Devonian shale an important part of the formation evaluation is to determine the amount of kerogen, since this appears as hydrocarbon-filled porosity to conventional logs. In this study Total Organic Carbon and pyrolysis analyses were made on 93 core samples from four wells. Based on these data a new method was used to drive volumetric kerogen and free oil content, and kerogen was found to range up to 26 percent by volume. A good correlation was subsequently developed to derive kerogen from the uranium response of the spectral gamma ray log. Another important result of this study is the measurement of formation water salinity directly on core samples. Results on 50 measurements in the four study wells ranged from 19,000 to 220,000 ppm NaCl

  12. Analysis of Doppler effect measurement in FCA cores using JENDL-3.2 library

    International Nuclear Information System (INIS)

    Okajima, Shigeaki

    1996-01-01

    For the evaluation of the calculation accuracy of the 238 U Doppler effect using JENDL-3.2 library, the previously measured Doppler reactivity worths in the FCA were systematically analyzed. In the analysis the Doppler reactivity worth was calculated by a first order perturbation theory. The calculated results were compared with those using JENDL-3.1 library. The JENDL-3.2 calculation in MOX fuel mock-up cores agrees well with the experimental values within the experimental error. In U-235/Pu fuel cores, the JENDL-3.2 calculation gives 12-15% larger Doppler reactivity worths than the JENDL-3.1 calculation. (author)

  13. Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

    DEFF Research Database (Denmark)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto

    2011-01-01

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features...

  14. Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

  15. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

    2016-12-15

    Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

  16. CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Park, Jong Woon

    2011-01-01

    The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

  17. Enterprise Architecture Modeling of Core Administrative Systems at KTH : A Modifiability Analysis

    OpenAIRE

    Rosell, Peter

    2012-01-01

    This project presents a case study of modifiability analysis on the Information Systems which are central to the core business processes of Royal Institution of Technology in Stockholm, Sweden by creating, updating and using models. The case study was limited to modifiability regarding only specified Information Systems. The method selected was Enterprise Architecture together with Enterprise Architecture Analysis research results and tools from the Industrial Information and Control Systems ...

  18. Comparative analysis of diagnostic accuracy of different brain biopsy procedures

    OpenAIRE

    Jain Deepali; Sharma Mehar; Sarkar Chitra; Gupta Deepak; Singh Manmohan; Mahapatra A

    2006-01-01

    Background: Image-guided procedures such as computed tomography (CT) guided, neuronavigator-guided and ultrasound-guided methods can assist neurosurgeons in localizing the intraparenchymal lesion of the brain. However, despite improvements in the imaging techniques, an accurate diagnosis of intrinsic lesion requires tissue sampling and histological verification. Aims: The present study was carried out to examine the reliability of the diagnoses made on tumor sample obtained via different s...

  19. Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

  20. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  1. Regional overpower protection system analysis for a DUPIC fuel CANDU core

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok; Park, Jee Won

    2003-06-01

    The regional overpower protection (ROP) system was assessed a CANDU 6 reactor with the DUPIC fuel, including the validation of the WIMS/RFSP/ROVER-F code system used for the estimation of ROP trip setpoint. The validation calculation has shown that it is valid to use the WIMS/RFSP/ROVER-F code system for ROP system analysis of the CANDU 6 core. For the DUPIC core, the ROP trip setpoint was estimated to be 125.7%, which is almost the same as that of the standard natural uranium core. This study has shown that the DUPIC fuel does not hurt the current ROP trip setpoint designed for the natural uranium CANDU 6 reactor

  2. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  3. Uncertainty analysis for the BEACON-COLSS core monitoring system application

    International Nuclear Information System (INIS)

    Morita, T.; Boyd, W.A.; Seong, K.B.

    2005-01-01

    This paper will cover the measurement uncertainty analysis of BEACON-COLSS core monitoring system. The uncertainty evaluation is made by using a BEACON-COLSS simulation program. By simulating the BEACON on-line operation for analytically generated reactor conditions, accuracy of the 'Measured' results can be evaluated by comparing to analytically generated 'Truth'. The DNB power margin is evaluated based on the Combustion Engineering's Modified Statistical Combination of Uncertainties (MSCU) using the CETOPD code for the DNBR calculation. A BEACON-COLSS simulation program for the uncertainty evaluation function has been established for plant applications. Qualification work has been completed for two Combustion Engineering plants. Results of the BEACON-COLSS measured peaking factors and DNBR power margin are plant type dependent and are applicable to reload cores as long as the core geometry and detector layout are unchanged. (authors)

  4. CoreFlow: A computational platform for integration, analysis and modeling of complex biological data

    DEFF Research Database (Denmark)

    Pasculescu, Adrian; Schoof, Erwin; Creixell, Pau

    2014-01-01

    between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion......A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which...... provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts...

  5. DANDE-a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

  6. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

  7. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem

  8. Structural, evolutionary and genetic analysis of the histidine biosynthetic "core" in the genus Burkholderia.

    Science.gov (United States)

    Papaleo, Maria Cristiana; Russo, Edda; Fondi, Marco; Emiliani, Giovanni; Frandi, Antonio; Brilli, Matteo; Pastorelli, Roberta; Fani, Renato

    2009-12-01

    In this work a detailed analysis of the structure, the expression and the organization of his genes belonging to the core of histidine biosynthesis (hisBHAF) in 40 newly determined and 13 available sequences of Burkholderia strains was carried out. Data obtained revealed a strong conservation of the structure and organization of these genes through the entire genus. The phylogenetic analysis showed the monophyletic origin of this gene cluster and indicated that it did not undergo horizontal gene transfer events. The analysis of the intergenic regions, based on the substitution rate, entropy plot and bendability suggested the existence of a putative transcription promoter upstream of hisB, that was supported by the genetic analysis that showed that this cluster was able to complement Escherichia colihisA, hisB, and hisF mutations. Moreover, a preliminary transcriptional analysis and the analysis of microarray data revealed that the expression of the his core was constitutive. These findings are in agreement with the fact that the entire Burkholderiahis operon is heterogeneous, in that it contains "alien" genes apparently not involved in histidine biosynthesis. Besides, they also support the idea that the proteobacterial his operon was piece-wisely assembled, i.e. through accretion of smaller units containing only some of the genes (eventually together with their own promoters) involved in this biosynthetic route. The correlation existing between the structure, organization and regulation of his "core" genes and the function(s) they perform in cellular metabolism is discussed.

  9. Microbial Analysis of Australian Dry Lake Cores; Analogs For Biogeochemical Processes

    Science.gov (United States)

    Nguyen, A. V.; Baldridge, A. M.; Thomson, B. J.

    2014-12-01

    Lake Gilmore in Western Australia is an acidic ephemeral lake that is analogous to Martian geochemical processes represented by interbedded phyllosilicates and sulfates. These areas demonstrate remnants of a global-scale change on Mars during the late Noachian era from a neutral to alkaline pH to relatively lower pH in the Hesperian era that continues to persist today. The geochemistry of these areas could possibly be caused by small-scale changes such as microbial metabolism. Two approaches were used to determine the presence of microbes in the Australian dry lake cores: DNA analysis and lipid analysis. Detecting DNA or lipids in the cores will provide evidence of living or deceased organisms since they provide distinct markers for life. Basic DNA analysis consists of extraction, amplification through PCR, plasmid cloning, and DNA sequencing. Once the sequence of unknown DNA is known, an online program, BLAST, will be used to identify the microbes for further analysis. The lipid analysis approach consists of phospholipid fatty acid analysis that is done by Microbial ID, which will provide direct identification any microbes from the presence of lipids. Identified microbes are then compared to mineralogy results from the x-ray diffraction of the core samples to determine if the types of metabolic reactions are consistent with the variation in composition in these analog deposits. If so, it provides intriguing implications for the presence of life in similar Martian deposits.

  10. Multivariate analysis of heavy metal contamination using river sediment cores of Nankan River, northern Taiwan

    Science.gov (United States)

    Lee, An-Sheng; Lu, Wei-Li; Huang, Jyh-Jaan; Chang, Queenie; Wei, Kuo-Yen; Lin, Chin-Jung; Liou, Sofia Ya Hsuan

    2016-04-01

    Through the geology and climate characteristic in Taiwan, generally rivers carry a lot of suspended particles. After these particles settled, they become sediments which are good sorbent for heavy metals in river system. Consequently, sediments can be found recording contamination footprint at low flow energy region, such as estuary. Seven sediment cores were collected along Nankan River, northern Taiwan, which is seriously contaminated by factory, household and agriculture input. Physico-chemical properties of these cores were derived from Itrax-XRF Core Scanner and grain size analysis. In order to interpret these complex data matrices, the multivariate statistical techniques (cluster analysis, factor analysis and discriminant analysis) were introduced to this study. Through the statistical determination, the result indicates four types of sediment. One of them represents contamination event which shows high concentration of Cu, Zn, Pb, Ni and Fe, and low concentration of Si and Zr. Furthermore, three possible contamination sources of this type of sediment were revealed by Factor Analysis. The combination of sediment analysis and multivariate statistical techniques used provides new insights into the contamination depositional history of Nankan River and could be similarly applied to other river systems to determine the scale of anthropogenic contamination.

  11. VERONA V6.22 – An enhanced reactor analysis tool applied for continuous core parameter monitoring at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Végh, J., E-mail: janos.vegh@ec.europa.eu [Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Pós, I., E-mail: pos@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Horváth, Cs., E-mail: csaba.horvath@energia.mta.hu [Centre for Energy Research, Hungarian Academy of Sciences, H-1525 Budapest 114, P.O. Box 49 (Hungary); Kálya, Z., E-mail: kalyaz@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Parkó, T., E-mail: parkot@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Ignits, M., E-mail: ignits@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary)

    2015-10-15

    Between 2003 and 2007 the Hungarian Paks NPP performed a large modernization project to upgrade its VERONA core monitoring system. The modernization work resulted in a state-of-the-art system that was able to support the reactor thermal power increase to 108% by more accurate and more frequent core analysis. Details of the new system are given in Végh et al. (2008), the most important improvements were as follows: complete replacement of the hardware and the local area network; application of a new operating system and porting a large fraction of the original application software to the new environment; implementation of a new human-system interface; and last but not least, introduction of new reactor physics calculations. Basic novelty of the modernized core analysis was the introduction of an on-line core-follow module based on the standard Paks NPP core design code HELIOS/C-PORCA. New calculations also provided much finer spatial resolution, both in terms of axial node numbers and within the fuel assemblies. The new system was able to calculate the fuel applied during the first phase of power increase accurately, but it was not tailored to determine the effects of burnable absorbers as gadolinium. However, in the second phase of the power increase process the application of fuel assemblies containing three fuel rods with gadolinium content was intended (in order to optimize fuel economy), therefore off-line and on-line VERONA reactor physics models had to be further modified, to be able to handle the new fuel according to the accuracy requirements. In the present paper first a brief overview of the system version (V6.0) commissioned after the first modernization step is outlined; then details of the modified off-line and on-line reactor physics calculations are described. Validation results for new modules are treated extensively, in order to illustrate the extent and complexity of the V&V procedure associated with the development and licensing of the new

  12. Modal analysis and acoustic transmission through offset-core honeycomb sandwich panels

    Science.gov (United States)

    Mathias, Adam Dustin

    The work presented in this thesis is motivated by an earlier research that showed that double, offset-core honeycomb sandwich panels increased thermal resistance and, hence, decreased heat transfer through the panels. This result lead to the hypothesis that these panels could be used for acoustic insulation. Using commercial finite element modeling software, COMSOL Multiphysics, the acoustical properties, specifically the transmission loss across a variety of offset-core honeycomb sandwich panels, is studied for the case of a plane acoustic wave impacting the panel at normal incidence. The transmission loss results are compared with those of single-core honeycomb panels with the same cell sizes. The fundamental frequencies of the panels are also computed in an attempt to better understand the vibrational modes of these particular sandwich-structured panels. To ensure that the finite element analysis software is adequate for the task at hand, two relevant benchmark problems are solved and compared with theory. Results from these benchmark results compared well to those obtained from theory. Transmission loss results from the offset-core honeycomb sandwich panels show increased transmission loss, especially for large cell honeycombs when compared to single-core honeycomb panels.

  13. Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

    International Nuclear Information System (INIS)

    Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

    1993-01-01

    Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

  14. Neutronics analysis on mini test fuel in the RSG-GAS core

    International Nuclear Information System (INIS)

    Tukiran S; Tagor M Sembiring

    2016-01-01

    Research on UMo fuel for research reactor has been developed. The fuel of research reactor is uranium molybdenum low enrichment with high density. For supporting the development of fuel fabrication, an neutronic analysis of mini fuel plates in the RSG-GAS core was performed. The aim of analysis is to determine the numbers of fuel cycles in the core to know the maximum fuel burn-up. The mini fuel plates of U_7Mo-Al and U_6Zr-Al with densities of 7.0 gU/cc and 5.2 gU/cc, respectively, will be irradiated in the RSG-GAS core. The size of both fuels, namely 630 x 70.75 x 1.30 mm were inserted to the 3 plates of dummy fuel. Before the fuel will be irradiated in the core, a calculation for safety analysis from neutronics and thermal-hydraulics aspects were required. However, in this paper, it will be discussed safety analysis of the U_7Mo-Al and U_6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF codes. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. If it is compared, the power density of U_7Mo-Al mini fuel is bigger than U_6Zr-Al fuel. (author)

  15. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  16. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  17. A Quantitative Review of Functional Analysis Procedures in Public School Settings

    Science.gov (United States)

    Solnick, Mark D.; Ardoin, Scott P.

    2010-01-01

    Functional behavioral assessments can consist of indirect, descriptive and experimental procedures, such as a functional analysis. Although the research contains numerous examples demonstrating the effectiveness of functional analysis procedures, experimental conditions are often difficult to implement in classroom settings and analog conditions…

  18. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  19. Analysis of loss of coolant accident and emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

    1977-01-01

    In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

  20. Application of Looped Network Analysis Method to Core of Prismatic VHTR

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

    2016-01-01

    Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively as shown in Fig. 1. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Moreover, it is hard to cover whole cases corresponding to the various bypass gap distribution in the whole VHTR core. In order to solve this problem, in this study, the flow network analysis code, FastNet (Flow Analysis for Steady-state Network), was developed using the Looped Network Analysis Method. The applied method was validated by comparing with SNU VHTR multi-block experiment. A 3-demensional network modeling was conducted representing flow paths as flow resistances. Flow network analysis code, FastNet, was developed to evaluate the core bypass flow distribution by using looped network analysis method. Complex flow network could be solved simply by converting the non-linear momentum equation to the linearized equation. The FastNet code predicted the flow distribution of the SNU multi-block experiment accurately

  1. Measurement and analysis of reaction rate distributions of cores with spectrum shifter region

    International Nuclear Information System (INIS)

    Matsuura, Shigekazu; Shiroya, Seiji; Unesaki, Hironobu; Takeda, Toshikazu; Aizawa, Otohiko; Kanda, Keiji.

    1995-01-01

    A study for the neutronic characteristics of the spectrum-controlled neutron irradiation fields using various reflector materials was performed. Spectrum shifter regions were constructed in the upper reflector region of the solid moderated core (B-Core) of the Kyoto University Critical Assembly (KUCA). Beryllium, graphite and aluminum were selected as the loading materials for the spectrum shifter. Two tight-pitch lattice cores with different moderator-to-fuel volume ratio (V m /V f ) of 0.97 and 0.65 have been used. Axial reaction rate distributions of gold, nickel and indium wires were measured, and the spectrum index was defined as the Cd ratio of the gold wire and the ratio of gold reaction rate to nickel reaction rate. Using the conventional design calculation procedure, the experimental and calculated reaction rate and spectrum index show several disagreements. Detailed treatment of the neutron streaming effect, heterogeneous cell structure and depression factor are shown to be necessary for improving the agreement between experimental and calculated values. (author)

  2. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-11-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

  3. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  4. Sample handling and chemical procedures for efficacious trace analysis of urine by neutron activation analysis

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Rack, E.P.; Roman, F.R.

    1988-01-01

    Important for the determination of trace elements, ions, or compounds in urine by chemical neutron activation analysis is the optimization of sample handling, preirradiation chemistry, and radioassay procedures necessary for viable analysis. Each element, because of its natural abundance in the earth's crust and, hence, its potential for reagent and environmental contamination, requires specific procedures for storage, handling, and preirradiation chemistry. Radioassay techniques for radionuclides vary depending on their half-lives and decay characteristics. Described in this paper are optimized procedures for aluminum and selenium. While 28 Al (T 1/2 = 2.24 min) and 77m Se(T 1/2 = 17.4s) have short half-lives, their gamma-ray spectra are quite different. Aluminum-28 decays by a 1779-keV gamma and 77m Se by a 162-keV gamma. Unlike selenium, aluminum is a ubiquitous element in the environment requiring special handling to minimize contamination in all phases of its analytical determination

  5. Content analysis of resident evaluations of faculty anesthesiologists: supervision encompasses some attributes of the professionalism core competency.

    Science.gov (United States)

    Dexter, Franklin; Szeluga, Debra; Hindman, Bradley J

    2017-05-01

    Anesthesiology departments need an instrument with which to assess practicing anesthesiologists' professionalism. The purpose of this retrospective analysis of the content of a cohort of resident evaluations of faculty anesthesiologists was to investigate the relationship between a clinical supervision scale and the multiple attributes of professionalism. From July 1, 2013 to the present, our department has utilized the de Oliveira Filho unidimensional nine-item supervision scale to assess the quality of clinical supervision of residents provided by our anesthesiologists. The "cohort" we examined included all 13,664 resident evaluations of all faculty anesthesiologists from July 1, 2013 through December 31, 2015, including 1,387 accompanying comments. Words and phrases associated with the core competency of professionalism were obtained from previous studies, and the supervision scale was analyzed for the presence of these words and phrases. The supervision scale assesses some attributes of anesthesiologists' professionalism as well as patient care and procedural skills and interpersonal and communication skills. The comments that residents provided with the below-average supervision scores included attributes of professionalism, although numerous words and phrases related to professionalism were not present in any of the residents' comments. The de Oliveira Filho clinical supervision scale includes some attributes of anesthesiologists' professionalism. The core competency of professionalism, however, is multidimensional, and the supervision scale and/or residents' comments did not address many of the other established attributes of professionalism.

  6. The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

    2006-07-15

    This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

  7. A new method based on low background instrumental neutron activation analysis for major, trace and ultra-trace element determination in atmospheric mineral dust from polar ice cores

    Energy Technology Data Exchange (ETDEWEB)

    Baccolo, Giovanni, E-mail: giovanni.baccolo@mib.infn.it [Graduate School in Polar Sciences, University of Siena, Via Laterina 8, 53100, Siena (Italy); Department of Environmental Sciences, University of Milano-Bicocca, P.zza della Scienza 1, 20126, Milano (Italy); INFN, Section of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy); Clemenza, Massimiliano [INFN, Section of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy); Department of Physics, University of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy); Delmonte, Barbara [Department of Environmental Sciences, University of Milano-Bicocca, P.zza della Scienza 1, 20126, Milano (Italy); Maffezzoli, Niccolò [Centre for Ice and Climate, Niels Bohr Institute, Juliane Maries Vej, 30, 2100, Copenhagen (Denmark); Nastasi, Massimiliano; Previtali, Ezio [INFN, Section of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy); Department of Physics, University of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy); Prata, Michele; Salvini, Andrea [LENA, University of Pavia, Pavia (Italy); Maggi, Valter [Department of Environmental Sciences, University of Milano-Bicocca, P.zza della Scienza 1, 20126, Milano (Italy); INFN, Section of Milano-Bicocca, P.zza della Scienza 3, 20126, Milano (Italy)

    2016-05-30

    Dust found in polar ice core samples present extremely low concentrations, in addition the availability of such samples is usually strictly limited. For these reasons the chemical and physical analysis of polar ice cores is an analytical challenge. In this work a new method based on low background instrumental neutron activation analysis (LB-INAA) for the multi-elemental characterization of the insoluble fraction of dust from polar ice cores is presented. Thanks to an accurate selection of the most proper materials and procedures it was possible to reach unprecedented analytical performances, suitable for ice core analyses. The method was applied to Antarctic ice core samples. Five samples of atmospheric dust (μg size) from ice sections of the Antarctic Talos Dome ice core were prepared and analyzed. A set of 37 elements was quantified, spanning from all the major elements (Na, Mg, Al, Si, K, Ca, Ti, Mn and Fe) to trace ones, including 10 (La, Ce, Nd, Sm, Eu, Tb, Ho, Tm, Yb and Lu) of the 14 natural occurring lanthanides. The detection limits are in the range of 10{sup −13}–10{sup −6} g, improving previous results of 1–3 orders of magnitude depending on the element; uncertainties lies between 4% and 60%. - Highlights: • A new method based on neutron activation for the multi-elemental characterization of atmospheric dust entrapped in polar ice cores is proposed. • 37 elements were quantified in μg size dust samples with detection limits ranging from 10{sup −13} to 10{sup −6} g. • A low background approach and a clean analytical protocol improved INAA performances to unprecedented levels for multi-elemental analyses.

  8. Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

    International Nuclear Information System (INIS)

    Pérez-Valseca, A.-D.; Espinosa-Paredes, G.; François, J.L.; Vázquez Rodríguez, A.; Martín-del-Campo, C.

    2017-01-01

    Highlights: • Methodology based on Monte Carlo simulation. • Sensitivity analysis of Lead Fast Reactor (LFR). • Uncertainty and regression analysis of LFR. • 10% change in the core inlet flow, the response in thermal power change is 0.58%. • 2.5% change in the inlet lead temperature the response is 1.87% in power. - Abstract: The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

  9. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  10. CoreFlow: a computational platform for integration, analysis and modeling of complex biological data.

    Science.gov (United States)

    Pasculescu, Adrian; Schoof, Erwin M; Creixell, Pau; Zheng, Yong; Olhovsky, Marina; Tian, Ruijun; So, Jonathan; Vanderlaan, Rachel D; Pawson, Tony; Linding, Rune; Colwill, Karen

    2014-04-04

    A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts into project-specific pipelines, tracks interdependencies between related tasks, and enables the generation of summary reports as well as publication-quality images. As a result, the gap between experimental and computational components of a typical large-scale biology project is reduced, decreasing the time between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion, and modeling of multiple/selected reaction monitoring (MRM/SRM) results. CoreFlow was purposely designed as an environment for programmers to rapidly perform data analysis. These analyses are assembled into project-specific workflows that are readily shared with biologists to guide the next stages of experimentation. Its simple yet powerful interface provides a structure where scripts can be written and tested virtually simultaneously to shorten the life cycle of code development for a particular task. The scripts are exposed at every step so that a user can quickly see the relationships between the data, the assumptions that have been made, and the manipulations that have been performed. Since the scripts use commonly available programming languages, they can easily be

  11. Analysis of sodium-void experiments in ZPPR-3 modified Phase 3 core

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, T.

    1978-08-01

    An analysis is presented of a series of sodium-void reactivity measurements performed in assembly 3 of Zero Power Plutonium Reactor (ZPPR-3), a mockup of the US Demoplant. In this series, large-zone sodium-void effects were studied in detail in the presence of many singularities, namely, control rods (CRs) and control rod positions (CRPs). The Karlsruhe data-and-method have been applied to an analysis of these experiments, and the results are presented. The work is aimed at complementing the sodium-void reactivity analysis based on the SNEAK experiments, where it was difficult to simulate a large plutonium-core of a prototype fast breeder reactor.

  12. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  13. Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

    Science.gov (United States)

    Preti, Raffaella

    2016-01-01

    The increased separation efficiency provided by the new technology of column packed with core-shell particles in high-performance liquid chromatography (HPLC) has resulted in their widespread diffusion in several analytical fields: from pharmaceutical, biological, environmental, and toxicological. The present paper presents their most recent applications in food analysis. Their use has proved to be particularly advantageous for the determination of compounds at trace levels or when a large amount of samples must be analyzed fast using reliable and solvent-saving apparatus. The literature hereby described shows how the outstanding performances provided by core-shell particles column on a traditional HPLC instruments are comparable to those obtained with a costly UHPLC instrumentation, making this novel column a promising key tool in food analysis. PMID:27143972

  14. Radiometric dating of sediment core from waterwork reservoir Rozgrund and analysis of mercury concentration depth profile

    International Nuclear Information System (INIS)

    Vanek, M.

    2005-01-01

    Radioisotope dating of lake sediments combined with analysis of chemical properties of the sediment layers allow us to study the history of the human impact on nature. Undisturbed sediment layers in the core samples serve as chronicle database with information about lake ecosystem and surrounding environment in the time of deposition. A sediment core sample from the bottom of the water-work reservoir Rozgrund was collected and separated into 2 cm thick layers. Samples were analysed by HPGe spectrometry for anthropogenous Cs-137 activity. From identified peaks corresponding to nuclear tests and Chernobyl accident the sedimentation rate was calculated and the chronology of layers established. Sub-samples from each layer were prepared separately for the analysis of the Hg concentration by atomic absorption spectrometry. The results show very small variations in Hg concentrations and there is no significant trend present in the profile. (author)

  15. Vibration Finite Element Analysis of SC10 Dry-type Transformer Core

    Directory of Open Access Journals (Sweden)

    Gao Sheng Wei

    2014-06-01

    Full Text Available As the popularization and application of dry-type power transformer, its work when the vibration noise problem widely concerned, on the basis of time-varying electromagnetic field and structural mechanics equation, this paper established a finite element analysis model of dry-type transformer, through the electromagnetic field – Structural mechanics field – sound field more than physical field coupling calculation analysis, obtained in no load and the vibration modes of the core under different load and frequency. According to the transformer vibration mechanism, compared with the experimental data, verified the accuracy of the calculation results, as the core of how to provide the theory foundation and to reduce the noise of the experiment.

  16. Nonlinear seismic analysis of a reactor structure with impact between core components

    International Nuclear Information System (INIS)

    Hill, R.G.

    1975-01-01

    The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-mass beam model of the FFTF which includes small clearances between core components is used as a ''driver'' for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed. 6 references

  17. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  18. Tank 241-TX-113 rotary mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    McCain, D.J.

    1998-01-01

    This sampling and analysis plan (SAP) identities characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-TX-113 (TX-113). The Tank Characterization Technical Sampling Basis document identities Retrieval, Pretreatment and Immobilization as an issue that applies to tank TX-113. As a result, a 150 gram composite of solids shall be made and archived for that program. This tank is not on a Watch List

  19. Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

    OpenAIRE

    Preti, Raffaella

    2016-01-01

    The increased separation efficiency provided by the new technology of column packed with core-shell particles in high-performance liquid chromatography (HPLC) has resulted in their widespread diffusion in several analytical fields: from pharmaceutical, biological, environmental, and toxicological. The present paper presents their most recent applications in food analysis. Their use has proved to be particularly advantageous for the determination of compounds at trace levels or when a large am...

  20. Numerical analysis of sandwich beam with corrugated core under three-point bending

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbeck, Leszek [Poznan University of Technology, Institute of Mathematics Piotrowo Street No. 5, 60-965 Poznan (Poland); Grygorowicz, Magdalena; Paczos, Piotr [Poznan University of Technology, Institute of Applied Mechanics Jana Pawla IIStreet No. 24, 60-965 Poznan (Poland)

    2015-03-10

    The strength problem of sandwich beam with corrugated core under three-point bending is presented.The beam are made of steel and formed by three mutually orthogonal corrugated layers. The finite element analysis (FEA) of the sandwich beam is performed with the use of the FEM system - ABAQUS. The relationship between the applied load and deflection in three-point bending is considered.

  1. Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire

    Science.gov (United States)

    2016-11-01

    Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire E n g in e e r R e s e a rc h a n d...id, age of the concrete being evaluated and tests performed...4 3 Preface This study was conducted in support of the Air Force Civil Engineer Center (AFCEC) to assess concrete obtained from Pease

  2. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-01-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ''hot spots'' do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ''first principles''. Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates

  3. Analysis on High Temperature Aging Property of Self-brazing Aluminum Honeycomb Core at Middle Temperature

    Directory of Open Access Journals (Sweden)

    ZHAO Huan

    2016-11-01

    Full Text Available Tension-shear test was carried out on middle temperature self-brazing aluminum honeycomb cores after high temperature aging by micro mechanical test system, and the microstructure and component of the joints were observed and analyzed using scanning electron microscopy and energy dispersive spectroscopy to study the relationship between brazing seam microstructure, component and high temperature aging properties. Results show that the tensile-shear strength of aluminum honeycomb core joints brazed by 1060 aluminum foil and aluminum composite brazing plate after high temperature aging(200℃/12h, 200℃/24h, 200℃/36h is similar to that of as-welded joints, and the weak part of the joint is the base metal which is near the brazing joint. The observation and analysis of the aluminum honeycomb core microstructure and component show that the component of Zn, Sn at brazing seam is not much affected and no compound phase formed after high temperature aging; therefore, the main reason for good high temperature aging performance of self-brazing aluminum honeycomb core is that no obvious change of brazing seam microstructure and component occurs.

  4. Provenance of whitefish in the Gulf of Bothnia determined by elemental analysis of otolith cores

    Science.gov (United States)

    Lill, J.-O.; Finnäs, V.; Slotte, J. M. K.; Jokikokko, E.; Heimbrand, Y.; Hägerstrand, H.

    2018-02-01

    The strontium concentration in the core of otoliths was used to determine the provenance of whitefish found in the Gulf of Bothnia, Baltic Sea. To that end, a database of strontium concentration in fish otoliths representing different habitats (sea, river and fresh water) had to be built. Otoliths from juvenile whitefish were therefore collected from freshwater ponds at 5 hatcheries, from adult whitefish from 6 spawning sites at sea along the Finnish west coast, and from adult whitefish ascending to spawn in the Torne River, in total 67 otoliths. PIXE was applied to determine the elemental concentrations in these otoliths. While otoliths from the juveniles raised in the freshwater ponds showed low but varying strontium concentrations (194-1664 μg/g,), otoliths from sea-spawning fish showed high uniform strontium levels (3720-4333 μg/g). The otolith core analysis of whitefish from Torne River showed large variations in the strontium concentrations (1525-3650 μg/g). These otolith data form a database to be used for provenance studies of wild adult whitefish caught at sea. The applicability of the database was evaluated by analyzing the core of polished otoliths from 11 whitefish from a test site at sea in the Larsmo archipelago. Our results show that by analyzing strontium in the otolith core, we can differentiate between hatchery-origin and wild-origin whitefish, but not always between river and sea spawning whitefish.

  5. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Hahn, D.

    2001-01-01

    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)

  6. Analysis of forces on core structures during a loss-of-coolant accident. Final report

    International Nuclear Information System (INIS)

    Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

    1980-08-01

    There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

  7. Investigation on macroscopic cross section model for BWR pin-by-pin core analysis - 118

    International Nuclear Information System (INIS)

    Fujita, T.; Tada, K.; Yamamoto, A.; Yamane, Y.; Kosaka, S.; Hirano, G.

    2010-01-01

    A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this effect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the reference cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points. (authors)

  8. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-01-01

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal's mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs

  9. A scenario-based procedure for seismic risk analysis

    International Nuclear Information System (INIS)

    Kluegel, J.-U.; Mualchin, L.; Panza, G.F.

    2006-12-01

    A new methodology for seismic risk analysis based on probabilistic interpretation of deterministic or scenario-based hazard analysis, in full compliance with the likelihood principle and therefore meeting the requirements of modern risk analysis, has been developed. The proposed methodology can easily be adjusted to deliver its output in a format required for safety analysts and civil engineers. The scenario-based approach allows the incorporation of all available information collected in a geological, seismotectonic and geotechnical database of the site of interest as well as advanced physical modelling techniques to provide a reliable and robust deterministic design basis for civil infrastructures. The robustness of this approach is of special importance for critical infrastructures. At the same time a scenario-based seismic hazard analysis allows the development of the required input for probabilistic risk assessment (PRA) as required by safety analysts and insurance companies. The scenario-based approach removes the ambiguity in the results of probabilistic seismic hazard analysis (PSHA) which relies on the projections of Gutenberg-Richter (G-R) equation. The problems in the validity of G-R projections, because of incomplete to total absence of data for making the projections, are still unresolved. Consequently, the information from G-R must not be used in decisions for design of critical structures or critical elements in a structure. The scenario-based methodology is strictly based on observable facts and data and complemented by physical modelling techniques, which can be submitted to a formalised validation process. By means of sensitivity analysis, knowledge gaps related to lack of data can be dealt with easily, due to the limited amount of scenarios to be investigated. The proposed seismic risk analysis can be used with confidence for planning, insurance and engineering applications. (author)

  10. Application of noise analysis for the study of core local instability at Forsmark 1

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-10-01

    Core local instability was recently experienced at Forsmark 1 BWR. The event has been studied by applying noise analysis to data collected in January 1997 for the stability test. The result indicated that there was a region in the left corner of the core which was subject to instability due to neutronic and thermal-hydraulic coupling. The result of the noise analysis suggested two types of disturbance source, one in the vicinity of the detector string LPRM10 having resonant oscillation at 0.5 Hz and another relatively wide band noise in the neighbourhood of LPRM18. Three hypotheses have been examined as the possible cause, operational factor, abnormal fuel assembly, and wide band low frequency disturbance. Although the real cause has not been made clear from the noise analysis, it is likely that the operational factor played an important role as the cause. Further investigations are expected to be performed in the future. In order to detect the local instability it is important to have a stability monitor with a capability of monitoring a sufficient number of LPRMs so as to cover the whole core. This is important since local instability is a type of anomaly which should not occur during reactor operation

  11. 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.

  12. A preliminary study on the behavior of trace elements in sediment cores from Ilha Grande (Rio de Janeiro State) by neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wasserman, Julio Cesar [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Dept. de Geoquimica; Figueiredo, Ana Maria G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Figueira, Andre Luiz [Universidade do Estado, Rio de Janeiro, RJ (Brazil). Dept. de Oceanografia; Kelecom, Alphonse [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Dept. de Biologia Geral

    2002-07-01

    The present work aims to identify atmospheric and marine inputs of 9 metals (Ba, Co, Cr, Cs, Fe, Hf, Rb, Sc, Zn), 8 rare earths (La, Ce, Nd, Sm, Eu, Tb, Yb e Lu), 2 actinides (U, Th) and 3 non-metals (As, Sb, Se) in sediment cores from a remote area, the Biological Reserve of Praia do Sul, Ilha Grande, Rio de Janeiro, Brazil. The sediment cores were sampled in a peat bog (out of the tidal reach) and in a mangrove, downstream of the peat bog. The analytical technique employed was Instrumental Neutron Activation Analysis. The samples were irradiated for 16 hours at a thermal neutron flux of 10{sup 12} n cm{sup -2} s{sup -1} at the IEA-R1 reactor of IPEN. The measurements of the induced gamma-ray activity were carried out by high resolution gamma spectrometry, with an hyperpure Ge detector. A preliminary sediment dating with Po-210 was also carried out by applying radiochemical procedures and measurements were done in an Alfa spectrometer The results indicate that the peat bog core present a slight surface enrichment that can be attributed to atmospheric inputs. Increasing concentrations of metals with age is probably due to history of soil occupation. In the mangrove core, no significant increase in concentration could be detected in the surface sediments (except for Zn) confirming the suitability of the peat bog core as a tracer for atmospheric inputs. (author)

  13. A preliminary study on the behavior of trace elements in sediment cores from Ilha Grande (Rio de Janeiro State) by neutron activation analysis

    International Nuclear Information System (INIS)

    Wasserman, Julio Cesar; Figueira, Andre Luiz; Kelecom, Alphonse

    2002-01-01

    The present work aims to identify atmospheric and marine inputs of 9 metals (Ba, Co, Cr, Cs, Fe, Hf, Rb, Sc, Zn), 8 rare earths (La, Ce, Nd, Sm, Eu, Tb, Yb e Lu), 2 actinides (U, Th) and 3 non-metals (As, Sb, Se) in sediment cores from a remote area, the Biological Reserve of Praia do Sul, Ilha Grande, Rio de Janeiro, Brazil. The sediment cores were sampled in a peat bog (out of the tidal reach) and in a mangrove, downstream of the peat bog. The analytical technique employed was Instrumental Neutron Activation Analysis. The samples were irradiated for 16 hours at a thermal neutron flux of 10 12 n cm -2 s -1 at the IEA-R1 reactor of IPEN. The measurements of the induced gamma-ray activity were carried out by high resolution gamma spectrometry, with an hyperpure Ge detector. A preliminary sediment dating with Po-210 was also carried out by applying radiochemical procedures and measurements were done in an Alfa spectrometer The results indicate that the peat bog core present a slight surface enrichment that can be attributed to atmospheric inputs. Increasing concentrations of metals with age is probably due to history of soil occupation. In the mangrove core, no significant increase in concentration could be detected in the surface sediments (except for Zn) confirming the suitability of the peat bog core as a tracer for atmospheric inputs. (author)

  14. INSIGHT: an integrated scoping analysis tool for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Noda, Hidefumi; Ito, Nobuaki; Maruyama, Taiji.

    1997-01-01

    An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis. (author)

  15. PROCEDURE FOR ANALYSIS AND EVALUATION OF MARKET POSITION PRODUCTION ORGANIZATION

    Directory of Open Access Journals (Sweden)

    A. N. Polozova

    2014-01-01

    Full Text Available Summary. Methodical procedures economic monitoring market position of industrial organization, particularly those relating to food production, including the 5 elements: matrix «component of business processes», matrix «materiality – efficiency», matrix «materiality – relevant», matrix emption and hindering factors matrix operation scenarios. Substantiated components assess the strengths and weaknesses of the business activities of organizations that characterize the state of internal business environment on the elements: production, organization, personnel, finance, marketing. The advantages of the matrix «materiality – relevance» consisting of 2 materiality level - high and low, and 3 directions relevance – «no change», «gain importance in the future», «lose importance in the future». Presented the contents of the matrix «scenarios functioning of the organization», involving 6 attribute levels, 10 classes of scenarios, 19 activities, including an optimistic and pessimistic. The evaluation of primary classes of scenarios, characterized by the properties of «development», «dynamic equilibrium», «quality improvement», «competitiveness», «favorable realization of opportunities», «competition resistance».

  16. Analysis of generalized Schwarz alternating procedure for domain decomposition

    Energy Technology Data Exchange (ETDEWEB)

    Engquist, B.; Zhao, Hongkai [Univ. of California, Los Angeles, CA (United States)

    1996-12-31

    The Schwartz alternating method(SAM) is the theoretical basis for domain decomposition which itself is a powerful tool both for parallel computation and for computing in complicated domains. The convergence rate of the classical SAM is very sensitive to the overlapping size between each subdomain, which is not desirable for most applications. We propose a generalized SAM procedure which is an extension of the modified SAM proposed by P.-L. Lions. Instead of using only Dirichlet data at the artificial boundary between subdomains, we take a convex combination of u and {partial_derivative}u/{partial_derivative}n, i.e. {partial_derivative}u/{partial_derivative}n + {Lambda}u, where {Lambda} is some {open_quotes}positive{close_quotes} operator. Convergence of the modified SAM without overlapping in a quite general setting has been proven by P.-L.Lions using delicate energy estimates. The important questions remain for the generalized SAM. (1) What is the most essential mechanism for convergence without overlapping? (2) Given the partial differential equation, what is the best choice for the positive operator {Lambda}? (3) In the overlapping case, is the generalized SAM superior to the classical SAM? (4) What is the convergence rate and what does it depend on? (5) Numerically can we obtain an easy to implement operator {Lambda} such that the convergence is independent of the mesh size. To analyze the convergence of the generalized SAM we focus, for simplicity, on the Poisson equation for two typical geometry in two subdomain case.

  17. Procedure for conducting a human-reliability analysis for nuclear power plants. Final report

    International Nuclear Information System (INIS)

    Bell, B.J.; Swain, A.D.

    1983-05-01

    This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study

  18. Failure mode and effects analysis: an empirical comparison of failure mode scoring procedures.

    Science.gov (United States)

    Ashley, Laura; Armitage, Gerry

    2010-12-01

    To empirically compare 2 different commonly used failure mode and effects analysis (FMEA) scoring procedures with respect to their resultant failure mode scores and prioritization: a mathematical procedure, where scores are assigned independently by FMEA team members and averaged, and a consensus procedure, where scores are agreed on by the FMEA team via discussion. A multidisciplinary team undertook a Healthcare FMEA of chemotherapy administration. This included mapping the chemotherapy process, identifying and scoring failure modes (potential errors) for each process step, and generating remedial strategies to counteract them. Failure modes were scored using both an independent mathematical procedure and a team consensus procedure. Almost three-fifths of the 30 failure modes generated were scored differently by the 2 procedures, and for just more than one-third of cases, the score discrepancy was substantial. Using the Healthcare FMEA prioritization cutoff score, almost twice as many failure modes were prioritized by the consensus procedure than by the mathematical procedure. This is the first study to empirically demonstrate that different FMEA scoring procedures can score and prioritize failure modes differently. It found considerable variability in individual team members' opinions on scores, which highlights the subjective and qualitative nature of failure mode scoring. A consensus scoring procedure may be most appropriate for FMEA as it allows variability in individuals' scores and rationales to become apparent and to be discussed and resolved by the team. It may also yield team learning and communication benefits unlikely to result from a mathematical procedure.

  19. Analysis of Relational Communication in Dyads: New Measurement Procedures.

    Science.gov (United States)

    Rogers, L. Edna; Farace, Richard

    Relational communication refers to the control or dominance aspects of message exchange in dyads--distinguishing it from the report or referential aspects of communication. In relational communicational analysis, messages as transactions are emphasized; major theoretical concepts which emerge are symmetry, transitoriness, and complementarity of…

  20. 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.

  1. 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.

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

    International Nuclear Information System (INIS)

    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 UZrH 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 ® 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

  3. Development of advanced nuclear core analysis system applicable to various reactor types

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

  4. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    International Nuclear Information System (INIS)

    Brunner, Andreas H; Sagmeister, Thomas; Kremer, Jolanta; Riss, Paul; Brustmann, Hermann

    2009-01-01

    Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB) in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. In our outpatient clinic120 CNB were performed. In 59/120 (49.2%) cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2%) were proved to be malignant and 17/59 (28.8%) to be benign in the definitive histology. 2/59 (3.3%) biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV) and negative predicting value (NPV) of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor

  5. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  6. A Bayesian multidimensional scaling procedure for the spatial analysis of revealed choice data

    NARCIS (Netherlands)

    DeSarbo, WS; Kim, Y; Fong, D

    1999-01-01

    We present a new Bayesian formulation of a vector multidimensional scaling procedure for the spatial analysis of binary choice data. The Gibbs sampler is gainfully employed to estimate the posterior distribution of the specified scalar products, bilinear model parameters. The computational procedure

  7. Responding to Self-Harm: A Documentary Analysis of Agency Policy and Procedure

    Science.gov (United States)

    Paul, Sally; Hill, Malcolm

    2013-01-01

    This paper reports on the findings of a documentary analysis of policies and procedures relating to self-harm from a range of organisations working with young people in the UK. It identifies the extent to which policies and/or procedures relating to self-harm are available for service providers and offers a wider understanding of the concepts of…

  8. False-negative results of breast core needle biopsies – retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Boba, Marek; Kołtun, Urszula; Bobek-Billewicz, Barbara; Chmielik, Ewa; Eksner, Bartosz; Olejnik, Tomasz

    2011-01-01

    Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. 988 core needle breast biopsies were performed at the Maria Skłodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration

  9. False-negative results of breast core needle biopsies - retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Boba, M.; Koltun, U.; Bobek-Billewicz, B.; Eksner, B.; Olejnik, T.; Chmielik, E.

    2011-01-01

    Background: Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. Material/Methods: 988 core needle breast biopsies were performed at the Maria Sklodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Results: Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Conclusions: Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration. (authors)

  10. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-03-01

    JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

  11. Subchannel analysis of a small ultra-long cycle fast reactor core

    International Nuclear Information System (INIS)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol

    2014-01-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria

  12. Fast analysis procedure of radiochemical coordinat uptake for methotrexate

    International Nuclear Information System (INIS)

    Caston, J.D.; Kamen, B.A.

    1976-01-01

    Under this invention, a radio-chemical analysis is submitted to determine the concentration of methotrexate or its equivalents in analysis in a biological medium. The amounts taken up of the labelled compound and the known concentrations of the unlabelled compound to be determined are radio-isotopically related to a first system containing a pre-determined amount of the labelled compound and a pre-determined amount of the unlabelled compound. In a second system, identical to the first, save that the sample of the biological medium to be analyzed takes the place of the unlabelled compound, the amount of labelled compound taken up is determined radio-isotopically. The concentration of the compound in the sample is then determined by correlation of the labelled compound uptake determined in the second system with the relation determined in the first system. The radio-isotopic relations and determinations may be made by direct and sequential analytical techniques [fr

  13. Development of the quantification procedures for in situ XRF analysis

    International Nuclear Information System (INIS)

    Kump, P.; Necemer, M.; Rupnik, P.

    2005-01-01

    For in situ XRF applications, two excitation systems (radioisotope and tube excited) and an X ray spectrometer based on an Si-PIN detector were assembled and used. The radioisotope excitation system with an Am-241 source was assembled into a prototype of a compact XRF analyser PEDUZO-01, which is also applicable in field work. The existing quantification software QAES (quantitative analysis of environmental samples) was assessed to be adequate also in field work. This QAES software was also integrated into a new software attached to the developed XRF analyser PEDUZO-01, which includes spectrum acquisition, spectrum analysis and quantification and runs in the LABVIEW environment. In a process of assessment of the Si-PIN based X ray spectrometers and QAES quantification software in field work, a comparison was made with the results obtained by the standard Si(Li) based spectrometer. The results of this study prove that the use of this spectrometer is adequate for field work. This work was accepted for publication in X ray Spectrometry. Application of a simple sample preparation of solid samples was studied in view of the analytical results obtained. It has been established that under definite conditions the results are not very different from the ones obtained by the homogenized sample pressed into the pellet. The influence of particle size and mineralogical effects on quantitative results was studied. A simple sample preparation kit was proposed. Sample preparation for the analysis of water samples by precipitation with APDC and aerosol analysis using a dichotomous sampler were also adapted and used in the field work. An adequate sample preparation kit was proposed. (author)

  14. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Mitchell, Jonathan; Fordham, Edmund J.

    2014-01-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  15. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    International Nuclear Information System (INIS)

    Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

    1986-10-01

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis

  16. Two-dimensional vertical model seismic test and analysis for HTGR core

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1983-02-01

    The resistance against earthquakes of high-temperature gas cooled reactor (HTGR) core with block-type fuels is not fully ascertained yet. Seismic studies must be made if such a reactor plant is to be installed in areas with frequent earthquakes. In the paper the test results of seismic behavior of a half-scale two-dimensional vertical slice core model and analysis are presented. The following results were obtained: (1) With soft spring support of the fixed side reflector structure, the relative column displacement is larger than that for hand support but the impact reaction force is smaller. (2) In the case of hard spring support the dowel force is smaller than for soft support. (3) The relative column displacement is larger in the core center than at the periphery. The impact acceleration (force) in the center is smaller than at the periphery. (4) The relative column displacement and impact reaction force are smaller with the gas pressure simulation spring than without. (5) With decreasing gap width between the top blocks of columns, the relative column displacement and impact reaction force decrease. (6) The column damping ratio was estimated as 4 -- 10% of critical. (7) The maximum impact reaction force for random waves such as seismic was below 60% that for a sinusoidal wave. (8) Vibration behavior and impact response are in good agreement between test and analysis. (author)

  17. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  18. Data management system for full core LOCA-analysis using TRANSURANUS

    International Nuclear Information System (INIS)

    Maertens, D.; Spykman, G.

    2005-01-01

    A data management system has been developed to perform full core pin by pin calculations of normal operation and (LOCA-) transient behaviour of fuel rods. The system automatically generates the input from a data base, controls the fuel rod calculations and provides a powerful tool for visualising the results. The full core pin by pin analysis now allows to use specific power histories, rod geometries and material data as well as enveloping data. Fuel rod code Transuranus is used for the normal operation and the transient phase in one run, thus assuring that the calculated rod properties of the normal operation (pre-transient) phase are handed over in all detail and not compressed to the transient phase. Transuranus has been upgraded with respect to high temperature models for Zry and M5 TM -cladding for creep, oxidation, heat rate dependent phase transition and anisotropy in the α and the mixed crystal phase. Parameter studies have been carried out to investigate the influence of using rod specific power histories instead of enveloping power histories in a full core analysis. The results show a significant increase in the ratio of failed fuel rods during a LOCA transient from 0.12% to approx. 50%. Another study for a typical PWR LOCA transient shows very good correlation between the distribution of failed fuel rods and rods with significant ballooning. (author)

  19. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  20. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

  1. A comparative examination of sample treatment procedures for ICAP-AES analysis of biological tissue

    Science.gov (United States)

    De Boer, J. L. M.; Maessen, F. J. M. J.

    The objective of this study was to contribute to the evaluation of existing sample preparation procedures for ICAP-AES analysis of biological material. Performance characteristics were established of current digestion procedures comprising extraction, solubilization, pressure digestion, and wet and dry ashing methods. Apart from accuracy and precision, a number of criteria of special interest for the analytical practice was applied. As a test sample served SRM bovine liver. In this material six elements were simultaneously determined. Results showed that every procedure has its defects and advantages. Hence, unambiguous recommendation of standard digestion procedures can be made only when taking into account the specific analytical problem.

  2. A comparison of various procedures in photon activation analysis with the same irradiation setup

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Z.J. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Wells, D. [Physics Department, South Dakota School of Mines and Technology, 501 E. Saint Joseph St., Rapid City, SD 57701 (United States); Segebade, C. [Idaho Accelerator Center, Idaho State University, 921 S. 8th Ave., Pocatello, ID 83209 (United States); Quigley, K.; Chemerisov, S. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2014-11-15

    A sample of known elemental concentrations was activated in the bremsstrahlung photon beam which was created by a pulsed electron LINAC. Several procedures of photon activation analysis, including those applied with/without reference material and with/without photon flux monitor, were conducted to make a comparison of their precision and accuracy in practice. Experimental results have indicated that: (1) relative procedures usually produce better outcome despite that the absolute measurement is straightforward and eliminate the assistance of reference materials; (2) among relative procedures, the method with internal flux monitor yields higher quality of the analytical results. In the article, the pros and cons of each procedure are discussed as well.

  3. Calculation and analysis of burnup and optimum core design in accelerator driven sub-critical system

    International Nuclear Information System (INIS)

    Wang Yuwei; Yang Yongwei; Cui Pengfei

    2011-01-01

    The premise of the accelerator driven sub-critical system (ADS) in the accident is still subcritical, the biggest k eff change with burn time is less than 1.5% and the cladding material, HT9 steel, can withstand the maximum radiation damage, core fuel area is divided into fuel transmutation area and fuel multiplication area, and fuel transmutation area maintains the same fuel composition in the whole process. Through the analysis of the composition of the fuel, shape of core layout and the power distribution, etc., supposed outer and inner Pu enrichment ratio range of 1.0-1.5, then the fuel components of fuel multiplication area was adjusted. Time evolution of k eff was calculated by COUPLED2 which coupled with MCNP and ORIGEN. At the same time the power peaking factors, minoractinides transmutation rate desired to maximization and burnup were considered. A sub-critical system fitting for engineering practice was established. (authors)

  4. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    International Nuclear Information System (INIS)

    Prabhu Gaunkar, N.; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C.; Bulu, I.; Ganesan, K.; Song, Y. Q.

    2015-01-01

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors

  5. SACI - O: A code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Resende Lobo, A.A. de; Soares, P.A.

    1979-03-01

    The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt

  6. Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan; FINAL

    International Nuclear Information System (INIS)

    TEMPLETON, A.M.

    1999-01-01

    This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AZ-102. Push mode core samples will be obtained from risers 15C and 24A to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples, composite the liquids and solids, perform chemical analyses, and provide subsamples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AZ-102 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plan

  7. Monte Carlo neutronics analysis of the ANS reactor three-element core design

    International Nuclear Information System (INIS)

    Wemple, C.A.

    1995-01-01

    The advanced neutron source (ANS) is a world-class research reactor and experimental center for neutron research, currently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(fission) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. It was designed to be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. This paper summarizes the neutronics efforts at the Idaho National Engineering Laboratory in support of the development and analysis of the three-element core for the advanced conceptual design phase

  8. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  9. Neutronic analysis of the Three Mile Island Unit 2 ex-core detector response

    International Nuclear Information System (INIS)

    Malloy, D.J.; Chang, Y.I.

    1981-10-01

    A neutronic analysis has been made with respect to the ex-core neutron detector response during the TMI-2 incident. A series of transport theory calculations quantified the impact upon the detector count rate of various core and downcomer conditions. In particular, various combinations of coolant void content and spatial distributions were investigated to yield the resulting transmission of the photoneutron source to the detector. The impact of a hypothetical distributed source within the downcomer region was also examined in order to simulate the potential effect of the release of neutron producing fission products into the coolant. These results are then offered as potential explanations for the anomalous behavior of the detector during the period of approx. 20 minutes through approx. 3 hours following the reactor scram

  10. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    International Nuclear Information System (INIS)

    Beard, Ch.; Morita, T.; Brown, J.

    2007-01-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  11. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Ch.; Morita, T.; Brown, J. [Westinghouse Electric Company, LLC, Nuclear Fuel Div., Pittsburgh, PA (United States)

    2007-07-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  12. Stereotactic core needle breast biopsy marker migration: An analysis of factors contributing to immediate marker migration.

    Science.gov (United States)

    Jain, Ashali; Khalid, Maria; Qureshi, Muhammad M; Georgian-Smith, Dianne; Kaplan, Jonah A; Buch, Karen; Grinstaff, Mark W; Hirsch, Ariel E; Hines, Neely L; Anderson, Stephan W; Gallagher, Katherine M; Bates, David D B; Bloch, B Nicolas

    2017-11-01

    To evaluate breast biopsy marker migration in stereotactic core needle biopsy procedures and identify contributing factors. This retrospective study analyzed 268 stereotactic biopsy markers placed in 263 consecutive patients undergoing stereotactic biopsies using 9G vacuum-assisted devices from August 2010-July 2013. Mammograms were reviewed and factors contributing to marker migration were evaluated. Basic descriptive statistics were calculated and comparisons were performed based on radiographically-confirmed marker migration. Of the 268 placed stereotactic biopsy markers, 35 (13.1%) migrated ≥1 cm from their biopsy cavity. Range: 1-6 cm; mean (± SD): 2.35 ± 1.22 cm. Of the 35 migrated biopsy markers, 9 (25.7%) migrated ≥3.5 cm. Patient age, biopsy pathology, number of cores, and left versus right breast were not associated with migration status (P> 0.10). Global fatty breast density (P= 0.025) and biopsy in the inner region of breast (P = 0.031) were associated with marker migration. Superior biopsy approach (P= 0.025), locally heterogeneous breast density, and t-shaped biopsy markers (P= 0.035) were significant for no marker migration. Multiple factors were found to influence marker migration. An overall migration rate of 13% supports endeavors of research groups actively developing new biopsy marker designs for improved resistance to migration. • Breast biopsy marker migration is documented in 13% of 268 procedures. • Marker migration is affected by physical, biological, and pathological factors. • Breast density, marker shape, needle approach etc. affect migration. • Study demonstrates marker migration prevalence; marker design improvements are needed.

  13. Using plant procedures as the basis for conducting a job and task analysis

    International Nuclear Information System (INIS)

    Haynes, F.H.; Ruth, B.W.

    1985-01-01

    Plant procedures were selected, by Northeast Utilities (NU), as the basis for conducting Job and Task Analyses (JTA). The resultant JTA was used to design procedure based simulator training programs for Millstone 1, 2, and Connecticut Yankee. The task listings were both plant specific and exhibited excellent correlation to INPO's generic PWR and BWR task analyses. Using the procedures based method enabled us to perform the JTA using plant and training staff. This proved cost effective in terms of both time and money. Learning objectives developed from the JTA were easily justified and correlated directly to job performance within the context of the plant procedures. In addition, the analysis generated a comprehensive review of plant procedures and, conversely, the plant's normal procedure revision process generated an automatic trigger for updating the task data

  14. The CERN antiproton target: hydrocode analysis of its core material dynamic response under proton beam impact

    CERN Document Server

    Martin, Claudio Torregrosa; Calviani, Marco; Muñoz-Cobo, José-Luis

    2016-01-01

    Antiprotons are produced at CERN by colliding a 26 GeV/c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 {\\deg}C after each pulse impact and successive dynamic pressure waves of the order of GPa's. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i) the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii) The existence of...

  15. CERN antiproton target: Hydrocode analysis of its core material dynamic response under proton beam impact

    Directory of Open Access Journals (Sweden)

    Claudio Torregrosa Martin

    2016-07-01

    Full Text Available Antiprotons are produced at CERN by colliding a 26  GeV/c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 °C after each pulse impact and successive dynamic pressure waves of the order of GPa’s. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii The existence of end-of-pulse tensile waves and its relevance on the overall response (iii A reduction of 44% in tensile pressure could be obtained by the use of a high density tantalum cladding.

  16. Safety analysis for push-mode and rotary-mode core sampling

    International Nuclear Information System (INIS)

    Milliken, N.J.; Geschke, G.R.

    1995-01-01

    This safety analysis analyzes using the push-mode core sampling truck in the push-mode and the rotary-mode core sampling trucks in both the push- and rotary-modes to retrieve core samples that, once taken and analyzed, will yield waste characterization data for the hazardous waste tanks at the Hanford Site. Operation of the core sampling trucks in both the push- and rotary-modes was reviewed to determine whether the release of radioactive materials could occur during operation. It was concluded that there are three credible scenarios: a sample spill outside of the tank, a steam release event, and an unfiltered release to the environment during continuous exhauster operation. The probability of a sample spill was found to be 10 -4 /event, the probability of a steam release event was determined to fall in the unlikely range (10 -2 /event to 10 -4 /event), and the probability of an unfiltered release was calculated to be 5 x 10 -3 /year. Typically, events with probabilities of 10 -6 /event or less are not considered to be risk significant, and the consequences usually are not analyzed. The three accident scenarios were analyzed to calculate the dose consequences. It was determined that the steam release event is the bounding accident. The onsite and offsite dose consequences for this event are calculated to be 0.24 Sv (24 rem) and 3.2 x 10 -4 Sv (32 mrem), respectively. These consequences are below the risk acceptance guidelines for an unlikely event, as established in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. With the design features and the use of the controls presented in Section 8.0, this operation represents a minimal risk

  17. Quality assurance procedures for the analysis of TRU waste samples

    International Nuclear Information System (INIS)

    Glasgow, D.C. Giaquinto, J.M.; Robinson, L.

    1995-01-01

    The Waste Isolation Pilot Plant (WIPP) project was undertaken in response to the growing need for a national repository for transuranic (TRU) waste. Guidelines for WIPP specify that any waste item to be interred must be fully characterized and analyzed to determine the presence of chemical compounds designated hazardous and certain toxic elements. The Transuranic Waste Characterization Program (TWCP) was launched to develop analysis and quality guidelines, certify laboratories, and to oversee the actual waste characterizations at the laboratories. ORNL is participating in the waste characterization phase and brings to bear a variety of analytical techniques including ICP-AES, cold vapor atomic absorption, and instrumental neutron activation analysis (INAA) to collective determine arsenic, cadmium, barium, chromium, mercury, selenium, silver, and other elements. All of the analytical techniques involved participate in a cooperative effort to meet the project objectives. One important component of any good quality assurance program is determining when an alternate method is more suitable for a given analytical problem. By bringing to bear a whole arsenal of analytical techniques working toward common objectives, few analytical problems prove to be insurmountable. INAA and ICP-AES form a powerful pair when functioning in this cooperative manner. This paper will provide details of the quality assurance protocols, typical results from quality control samples for both INAA and ICP-AES, and detail method cooperation schemes used

  18. Procedures for uncertainty and sensitivity analysis in repository performance assessment

    International Nuclear Information System (INIS)

    Poern, K.; Aakerlund, O.

    1985-10-01

    The objective of the project was mainly a literature study of available methods for the treatment of parameter uncertainty propagation and sensitivity aspects in complete models such as those concerning geologic disposal of radioactive waste. The study, which has run parallel with the development of a code package (PROPER) for computer assisted analysis of function, also aims at the choice of accurate, cost-affective methods for uncertainty and sensitivity analysis. Such a choice depends on several factors like the number of input parameters, the capacity of the model and the computer reresources required to use the model. Two basic approaches are addressed in the report. In one of these the model of interest is directly simulated by an efficient sampling technique to generate an output distribution. Applying the other basic method the model is replaced by an approximating analytical response surface, which is then used in the sampling phase or in moment matching to generate the output distribution. Both approaches are illustrated by simple examples in the report. (author)

  19. A finite element thermal analysis of various dowel and core materials

    Directory of Open Access Journals (Sweden)

    Shanti Varghese

    2012-01-01

    Conclusion: Non-metallic dowel and core materials such as fibre reinforced composite dowels (FRC generate greater stress than metallic dowel and core materials. This emphasized the preferable use of the metallic dowel and core materials in the oral environment.

  20. Analysis of core-periphery organization in protein contact networks reveals groups of structurally and functionally critical residues.

    Science.gov (United States)

    Isaac, Arnold Emerson; Sinha, Sitabhra

    2015-10-01

    The representation of proteins as networks of interacting amino acids, referred to as protein contact networks (PCN), and their subsequent analyses using graph theoretic tools, can provide novel insights into the key functional roles of specific groups of residues. We have characterized the networks corresponding to the native states of 66 proteins (belonging to different families) in terms of their core-periphery organization. The resulting hierarchical classification of the amino acid constituents of a protein arranges the residues into successive layers - having higher core order - with increasing connection density, ranging from a sparsely linked periphery to a densely intra-connected core (distinct from the earlier concept of protein core defined in terms of the three-dimensional geometry of the native state, which has least solvent accessibility). Our results show that residues in the inner cores are more conserved than those at the periphery. Underlining the functional importance of the network core, we see that the receptor sites for known ligand molecules of most proteins occur in the innermost core. Furthermore, the association of residues with structural pockets and cavities in binding or active sites increases with the core order. From mutation sensitivity analysis, we show that the probability of deleterious or intolerant mutations also increases with the core order. We also show that stabilization centre residues are in the innermost cores, suggesting that the network core is critically important in maintaining the structural stability of the protein. A publicly available Web resource for performing core-periphery analysis of any protein whose native state is known has been made available by us at http://www.imsc.res.in/ ~sitabhra/proteinKcore/index.html.

  1. Oxygen isotope analysis of plant water without extraction procedure

    International Nuclear Information System (INIS)

    Gan, K.S.; Wong, S.C.; Farquhar, G.D.; Yong, J.W.H.

    2001-01-01

    Isotopic analyses of plant water (mainly xylem, phloem and leaf water) are gaming importance as the isotopic signals reflect plant-environment interactions, affect the oxygen isotopic composition of atmospheric O 2 and CO 2 and are eventually incorporated into plant organic matter. Conventionally, such isotopic measurements require a time-consuming process of isolating the plant water by azeotropic distillation or vacuum extraction, which would not complement the speed of isotope analysis provided by continuous-flow IRMS (Isotope-Ratio Mass Spectrometry), especially when large data sets are needed for statistical calculations in biological studies. Further, a substantial amount of plant material is needed for water extraction and leaf samples would invariably include unenriched water from the fine veins. To measure sub-microlitre amount of leaf mesophyll water, a new approach is undertaken where a small disc of fresh leaf is cut using a specially designed leaf punch, and pyrolysed directly in an IRMS. By comparing with results from pyrolysis of the dry matter of the same leaf, the 18 O content of leaf water can be determined without extraction from fresh leaves. This method is validated using a range of cellulose-water mixtures to simulate the constituents of fresh leaf. Cotton leaf water δ 18 O obtained from both methods of fresh leaf pyrolysis and azeotropic distillation will be compared. The pyrolysis technique provides a robust approach to measure the isotopic content of water or any volatile present in a homogeneous solution or solid hydrous substance

  2. Input Files and Procedures for Analysis of SMA Hybrid Composite Beams in MSC.Nastran and ABAQUS

    Science.gov (United States)

    Turner, Travis L.; Patel, Hemant D.

    2005-01-01

    A thermoelastic constitutive model for shape memory alloys (SMAs) and SMA hybrid composites (SMAHCs) was recently implemented in the commercial codes MSC.Nastran and ABAQUS. The model is implemented and supported within the core of the commercial codes, so no user subroutines or external calculations are necessary. The model and resulting structural analysis has been previously demonstrated and experimentally verified for thermoelastic, vibration and acoustic, and structural shape control applications. The commercial implementations are described in related documents cited in the references, where various results are also shown that validate the commercial implementations relative to a research code. This paper is a companion to those documents in that it provides additional detail on the actual input files and solution procedures and serves as a repository for ASCII text versions of the input files necessary for duplication of the available results.

  3. Reliability and accuracy of a video analysis protocol to assess core ability.

    Science.gov (United States)

    McDonald, Dawn A; Delgadillo, James Q; Fredericson, Michael; McConnell, Jennifer; Hodgins, Melissa; Besier, Thor F

    2011-03-01

    To develop and test a method to measure core ability in healthy athletes with 2-dimensional video analysis software (SiliconCOACH). Specific objectives were to: (1) develop a standardized exercise battery with progressions of increasing difficulty to evaluate areas of core ability in elite athletes; (2) develop an objective and quantitative grading rubric with the use of video analysis software; (3) assess the test-retest reliability of the exercise battery; (4) assess the interrater and intrarater reliability of the video analysis system; and (5) assess the accuracy of the assessment. Test-retest repeatability and accuracy. Testing was conducted in the Stanford Human Performance Laboratory, Stanford University, Stanford, CA. Nine female gymnasts currently training with the Stanford Varsity Women's Gymnastics Team participated in testing. Participants completed a test battery composed of planks, side planks, and leg bridges of increasing difficulty. Subjects completed two 20-minute testing sessions within a 4- to 10-day period. Two-dimensional sagittal-plane video was captured simultaneously with 3-dimensional motion capture. The main outcome measures were pelvic displacement and time that elapsed until failure occurred, as measured with SiliconCOACH video analysis software. Test-retest and interrater and intrarater reliability of the video analysis measures was assessed. Accuracy as compared with 3-dimensional motion capture also was assessed. Levels reached during the side planks and leg bridges had an excellent test-retest correlation (r(2) = 0.84, r(2) = 0.95). Pelvis displacements measured by examiner 1 and examiner 2 had an excellent correlation (r(2) = 0.86, intraclass correlation coefficient = 0.92). Pelvis displacements measured by examiner 1 during independent grading sessions had an excellent correlation (r(2) = 0.92). Pelvis displacements from the plank and from a set of combined plank and side plank exercises both had an excellent correlation with 3

  4. Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2001-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

  5. Modeling analysis of pulsed magnetization process of magnetic core based on inverse Jiles-Atherton model

    Science.gov (United States)

    Liu, Yi; Zhang, He; Liu, Siwei; Lin, Fuchang

    2018-05-01

    The J-A (Jiles-Atherton) model is widely used to describe the magnetization characteristics of magnetic cores in a low-frequency alternating field. However, this model is deficient in the quantitative analysis of the eddy current loss and residual loss in a high-frequency magnetic field. Based on the decomposition of magnetization intensity, an inverse J-A model is established which uses magnetic flux density B as an input variable. Static and dynamic core losses under high frequency excitation are separated based on the inverse J-A model. Optimized parameters of the inverse J-A model are obtained based on particle swarm optimization. The platform for the pulsed magnetization characteristic test is designed and constructed. The hysteresis curves of ferrite and Fe-based nanocrystalline cores at high magnetization rates are measured. The simulated and measured hysteresis curves are presented and compared. It is found that the inverse J-A model can be used to describe the magnetization characteristics at high magnetization rates and to separate the static loss and dynamic loss accurately.

  6. Reference accident (Core disruption accident - safety analysis detailed report no. 11)

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The PEC safety analysis led to the conclusion that all credible sequences (incident sequences characterized by a frequency of occurrence above 10/sup minus 7/ events per year) are limited to the design basis conditions of components of the plant protection systems, and that none of them leads to a release of mechanical energy or to an extensive damage of the core and primary containment structures event in the case of failure to scram. Nevertheless, as is done in other countries for similar reactors, some events beyond the limits of credibility were considered for the PEC reactor. These were defined on a absolutely hypothetical basis that involves severe core disruption and dynamic loading of primary containment boundary. A series of containments, each having a different role, was designed to mitigate the radiological effects of a postulated core disruptive accident. The final aim was to demonstrate that residual heat can be removed and that the release of radioactivity to the environment is within acceptable limits.

  7. Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

    1978-07-01

    The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

  8. Development of spectral history methods for pin-by-pin core analysis method using three-dimensional direct response matrix

    International Nuclear Information System (INIS)

    Mitsuyasu, T.; Ishii, K.; Hino, T.; Aoyama, M.

    2009-01-01

    Spectral history methods for pin-by-pin core analysis method using the three-dimensional direct response matrix have been developed. The direct response matrix is formalized by four sub-response matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the burn-up effect related to spectral history. One of the methods is to evaluate the nodal burn-up spectrum obtained using the out-going neutron current. The other is to correct the fuel rod neutron production rates obtained the pin-by-pin correction. These spectral history methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error can be reduced by half during burn-up, the nodal neutron production rates errors can be reduced by 30% or more. The root-mean-square differences between the relative fuel rod neutron production rate distributions can be reduced within 1.1% error. This means that these methods can accurately reflect the effects of intra- and inter-assembly heterogeneities during burn-up and can be used for core analysis. Core analysis with the DRM method was carried out for an ABWR quarter core and it was found that both thermal power and coolant-flow distributions were smoothly converged. (authors)

  9. Third Generation (3G) Site Characterization: Cryogenic Core Collection and High Throughput Core Analysis - An Addendum to Basic Research Addressing Contaminants in Low Permeability Zones - A State of the Science Review

    Science.gov (United States)

    2016-07-29

    Styrofoam insulation for keeping the core frozen during MRI .................................. 78 Figure 5-2. Schematic of reference and core setting in... Hollow -Stem Auger HTCA High-Throughput Core Analysis IC Ion Chromatograph ID Inner Diameter k Permeability LN Liquid Nitrogen LNAPL Light...vibration, or “over drilling” using a hollow -stem auger. The ratio of the length of the collected core to the depth over which the sample tube is

  10. An analysis of tolerance levels in IMRT quality assurance procedures

    International Nuclear Information System (INIS)

    Basran, Parminder S.; Woo, Milton K.

    2008-01-01

    Increased use of intensity modulated radiation therapy (IMRT) has resulted in increased efforts in patient quality assurance (QA). Software and detector systems intended to streamline the IMRT quality assurance process often report metrics, such as percent discrepancies between measured and computed doses, which can be compared to benchmark or threshold values. The purpose of this work is to examine the relationships between two different types of IMRT QA processes in order to define, or refine, appropriate tolerances values. For 115 IMRT plans delivered in a 3 month period, we examine the discrepancies between (a) the treatment planning system (TPS) and results from a commercial independent monitor unit (MU) calculation program; (b) TPS and results from a commercial diode-array measurement system; and (c) the independent MU calculation and the diode-array measurements. Statistical tests were performed to assess significance in the IMRT QA results for different disease site and machine models. There is no evidence that the average total dose discrepancy in the monitor unit calculation depends on the disease site. Second, the discrepancies in the two IMRT QA methods are independent: there is no evidence that a better --or worse--monitor unit validation result is related to a better--or worse--diode-array measurement result. Third, there is marginal benefit in repeating the independent MU calculation with a more suitable dose point, if the initial IMRT QA failed a certain tolerance. Based on these findings, the authors conclude at some acceptable tolerances based on disease site and IMRT QA method. Specifically, monitor unit validations are expected to have a total dose discrepancy of 3% overall, and 5% per beam, independent of disease site. Diode array measurements are expected to have a total absolute dose discrepancy of 3% overall, and 3% per beam, independent of disease site. The percent of pixels exceeding a 3% and 3 mm threshold in a gamma analysis should be

  11. Analysis of Few-Mode Multi-Core Fiber Splice Behavior Using an Optical Vector Network Analyzer

    DEFF Research Database (Denmark)

    Rommel, Simon; Mendinueta, Jose Manuel Delgado; Klaus, Werner

    2017-01-01

    The behavior of splices in a 3-mode 36-core fiber is analyzed using optical vector network analysis. Time-domain response analysis confirms splices may cause significant mode-mixing, while frequency-domain analysis shows splices may affect system level mode-dependent loss both positively and negativ......The behavior of splices in a 3-mode 36-core fiber is analyzed using optical vector network analysis. Time-domain response analysis confirms splices may cause significant mode-mixing, while frequency-domain analysis shows splices may affect system level mode-dependent loss both positively...

  12. Few-Group Transport Analysis of the Core-Reflector Problem in Fast Reactor Cores via Equivalent Group Condensation and Local/Global Iteration

    International Nuclear Information System (INIS)

    Won, Jong Hyuck; Cho, Nam Zin

    2011-01-01

    In deterministic neutron transport methods, a process called fine-group to few-group condensation is used to reduce the computational burden. However, recent results on the core-reflector problem in fast reactor cores show that use of a small number of energy groups has limitation to describe neutron flux around core reflector interface. Therefore, researches are still ongoing to overcome this limitation. Recently, the authors proposed I) direct application of equivalently condensed angle-dependent total cross section to discrete ordinates method to overcome the limitation of conventional multi-group approximations, and II) local/global iteration framework in which fine-group discrete ordinates calculation is used in local problems while few-group transport calculation is used in the global problem iteratively. In this paper, an analysis of the core-reflector problem is performed in few-group structure using equivalent angle-dependent total cross section with local/global iteration. Numerical results are obtained under S 12 discrete ordinates-like transport method with scattering cross section up to P1 Legendre expansion

  13. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    Rebollo, L.; Blanco, J.

    2001-01-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  14. Microarray analysis in clinical oncology: pre-clinical optimization using needle core biopsies from xenograft tumors

    International Nuclear Information System (INIS)

    Goley, Elizabeth M; Anderson, Soni J; Ménard, Cynthia; Chuang, Eric; Lü, Xing; Tofilon, Philip J; Camphausen, Kevin

    2004-01-01

    labeled, would generate representative array profiles compared to larger excisional biopsy material. In this analysis correlation coefficients were obtained ranging from 0.750–0.834 between U251 biopsy cores and excised tumors, and 0.812–0.846 between DU145 biopsy cores and excised tumors. These data suggest that needle core biopsies can be used as reliable tissue samples for tumor microarray analysis after linear amplification and either indirect or direct labeling of the starting RNA

  15. Genetic diversity and population structure analysis to construct a core collection from a large Capsicum germplasm.

    Science.gov (United States)

    Lee, Hea-Young; Ro, Na-Young; Jeong, Hee-Jin; Kwon, Jin-Kyung; Jo, Jinkwan; Ha, Yeaseong; Jung, Ayoung; Han, Ji-Woong; Venkatesh, Jelli; Kang, Byoung-Cheorl

    2016-11-14

    Conservation of genetic diversity is an essential prerequisite for developing new cultivars with desirable agronomic traits. Although a large number of germplasm collections have been established worldwide, many of them face major difficulties due to large size and a lack of adequate information about population structure and genetic diversity. Core collection with a minimum number of accessions and maximum genetic diversity of pepper species and its wild relatives will facilitate easy access to genetic material as well as the use of hidden genetic diversity in Capsicum. To explore genetic diversity and population structure, we investigated patterns of molecular diversity using a transcriptome-based 48 single nucleotide polymorphisms (SNPs) in a large germplasm collection comprising 3,821 accessions. Among the 11 species examined, Capsicum annuum showed the highest genetic diversity (H E  = 0.44, I = 0.69), whereas the wild species C. galapagoense showed the lowest genetic diversity (H E  = 0.06, I = 0.07). The Capsicum germplasm collection was divided into 10 clusters (cluster 1 to 10) based on population structure analysis, and five groups (group A to E) based on phylogenetic analysis. Capsicum accessions from the five distinct groups in an unrooted phylogenetic tree showed taxonomic distinctness and reflected their geographic origins. Most of the accessions from European countries are distributed in the A and B groups, whereas the accessions from Asian countries are mainly distributed in C and D groups. Five different sampling strategies with diverse genetic clustering methods were used to select the optimal method for constructing the core collection. Using a number of allelic variations based on 48 SNP markers and 32 different phenotypic/morphological traits, a core collection 'CC240' with a total of 240 accessions (5.2 %) was selected from within the entire Capsicum germplasm. Compared to the other core collections, CC240 displayed higher

  16. Analysis of 2D reactor core using linear perturbation theory and nodal finite element methods

    International Nuclear Information System (INIS)

    Adrian Mugica; Edmundo del Valle

    2005-01-01

    In this work the multigroup steady state neutron diffusion equations are solved using the nodal finite element method (NFEM) and the Linear Perturbation Theory (LPT) for XY geometry. The NFEM used corresponds to the Raviart-Thomas schemes RT0 and RT1, interpolating 5 and 12 parameters respectively in each node of the space discretization. The accuracy of these methods is related with the dimension of the space approximation and the mesh size. Therefore, using fine meshes and the RT0 or RT1 nodal methods leads to a large an interesting eigenvalue problem. The finite element method used to discretize the weak formulation of the diffusion equations is the Galerkin one. The algebraic structure of the discrete eigenvalue problem is obtained and solved using the Wielandt technique and the BGSTAB iterative method using the SPARSKIT package developed by Yousef Saad. The results obtained with LPT show good agreement with the results obtained directly for the perturbed problem. In fact, the cpu time to solve a single problem, the unperturbed and the perturbed one, is practically the same but when one is focused in shuffling many times two different assemblies in the core then the LPT technique becomes quite useful to get good approximations in a short time. This particular problem was solved for one quarter-core with NFEM. Thus, the computer program based on LPT can be used to perform like an analysis tool in the fuel reload optimization or combinatory analysis to get reload patterns in nuclear power plants once that it had been incorporated with the thermohydraulic aspects needed to simulate accurately a real problem. The maximum differences between the NFEM and LPT for the three LWR reactor cores are about 250 pcm. This quantity is considered an acceptable value for this kind of analysis. (authors)

  17. A coupling model for the two-stage core calculation method with subchannel analysis for boiling water reactors

    International Nuclear Information System (INIS)

    Mitsuyasu, Takeshi; Aoyama, Motoo; Yamamoto, Akio

    2017-01-01

    Highlights: • A coupling model of the two-stage core calculation with subchannel analysis. • BWR fuel assembly parameters are assumed and verified. • The model was evaluated for heterogeneous problems. - Abstract: The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted cond