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Sample records for core analysis revision

  1. 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)

  2. 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)

  3. Avian Analysis of LCTA Core Plot Data: West Point Military Academy (Revised)

    National Research Council Canada - National Science Library

    Schreiber, Eric

    1998-01-01

    The Land Condition Trend Analysis (LCTA) program is the Army's standard for land inventory and monitoring, using standard methods for natural resources data collection, analyses, and reporting designed to meet multiple goals and objectives...

  4. Revised South China Sea spreading history based on macrostructure analysis of IODP Expedition 349 core samples and geophysical data

    Science.gov (United States)

    Sun, Z.; Ding, W.; Zhao, X.; Qiu, N.; Lin, J.; Li, C.

    2017-12-01

    In Internaltional Ocean Discovery Program (IODP) Expedition 349, four sites were drilled and cored successfully in the South China Sea (SCS). Three of them are close to the central spreading ridge (Sites U1431, U1433 and U1434), and one (Site U1435) is located on an outer rise,,providingsignificant information on the spreading history of the SCS.In order to constrain the spreading historymore accurately with the core results, we analyzed the identifiable macrostructures (over 300 fractures, veins and slickensides)from all the consolidated samples of these four drill sites. Then we made a retrograde reconstruction of the SCS spreading history with the constraints of the estimated fractures and veins, post-spreading volcanism,seismic interpretation, as well as free-air gravity and magnetic anomaly and topography analysis. Our study indicates that the spreading of the SCS experienced at least one ridge jump event and two events of ridge orientation and spreading direction adjustment, which mademagnetic anomaly orientation, ridge positionand facture zone directionskeep changing in the South China Sea. During the last spreading stage, the spreading direction was north-southward but lasted for a very short time period. The oceanic crust is wider in the eastern SCS and tapers out toward west.Due to the subductionof SCS beneath the Philippine Sea plate, the seafloor began to develop new fractures:the NWW-to EW-trending R' shear faults and the NE-trending P faultsbecame dominant faults and controlled the eruption of post-drift volcanism.

  5. Summary of micrographic analysis of fracture coating phases on drill cores from Pahute Mesa, Nevada Test Site. Revision 1

    International Nuclear Information System (INIS)

    1998-12-01

    The flow path between Pahute Mesa and the groundwater discharge area in Oasis Valley (approximately 18 miles to the southwest) is of concern due to the relatively short travel distance between a recharge area where underground nuclear testing has been conducted and the off-site water users. Groundwater flow and transport modeling by IT Corporation (IT) has shown rapid tritium transport in the volcanic rock aquifers along this flow path. The resultant estimates of rapid transport were based on water level data, limited hydraulic conductivity data, estimates of groundwater discharge rates in Oasis Valley, assumed porosities, and estimated retardation rates. Many of these parameters are poorly constrained and may vary considerably. Sampling and analytical techniques are being applied as an independent means to determine transport rates by providing an understanding of the geochemical processes that control solute movement along the flow path. As part of these geochemical investigations, this report summarizes the analysis of fracture coating mineral phases from drill core samples from the Pahute mesa area of the Nevada Test Site (NTS). Archived samples were collected based on the presence of natural fractures and on the types and abundance of secondary mineral phases present on those fracture surfaces. Mineral phases present along fracture surfaces are significant because, through the process of water-rock interaction, they can either contribute (as a result of dissolution) or remove (as a result of precipitation or adsorption) constituents from solution. Particular attention was paid to secondary calcite occurrences because they represent a potential source of exchangeable carbon and can interact with groundwater resulting in a modified isotopic signature and apparent water age

  6. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I

    International Nuclear Information System (INIS)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes

  7. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  8. 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.)

  9. 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)

  10. BWRVIP-123, Revision 1NP: BWR Vessel and Internals Project Removal and Analysis of Material Samples from Core Shroud and Top Guide at Susquehanna Unit 2

    International Nuclear Information System (INIS)

    Howell, D.; Haertel, T.; Lindberg, J.; Oliver, B.; Greenwood, L.

    2005-01-01

    Fast and thermal fluence were determined by a laboratory analysis of the samples. Fluence in the upper regions of the shroud (between the H1 and H2 welds) was substantially lower than that in the belt line region (near the H4 weld). Fluence in the top guide was significantly higher than fluence on the core shroud. As expected, helium concentrations were highest in regions where fluence was highest. Estimates of the initial boron concentration were similar to measurements made on materials removed from other reactors. A technical justification evaluated the acceptability of the sampling process with respect to structural consequences of material removal and to increased cracking susceptibility due to the as-left condition. It was determined that the sampling process was acceptable on both counts

  11. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I. [DEPTH-CHARGE code

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes.

  12. 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

  13. Revised

    DEFF Research Database (Denmark)

    Johannsen, Vivian Kvist; Nord-Larsen, Thomas; Riis-Nielsen, Torben

    This report is a revised analysis of the Danish data on CO2 emissions from forest, afforestation and deforestation for the period 1990 - 2008 and a prognosis for the period until 2020. Revision have included measurements from 2009 in the estimations. The report is funded by the Ministry of Climate...

  14. Program plan for shipment, receipt, and storage of the TMI-2 core. Revision 1

    International Nuclear Information System (INIS)

    Quinn, G.J.; Reno, H.W.; Schmitt, R.C.

    1985-01-01

    This plan addresses the preparation and shipment of core debris from Three Mile Island Unit 2 (TMI-2) to the Idaho National Engineering Laboratory (INEL) and receipt and storage of that core debris. The Manager of the Nuclear Materials Evaluation Programs Division of EG and G Idaho, Inc. will manage two separate but integrated programs, one located at TMI (Part 1) and the other at INEL (Part 2). The Technical Integration Office (at TMI) is responsible for developing and implementing Part 1, TMI-2 Core Shipment Program. The Technical Support Branch (at INEL) is responsible for developing and implementing Part 2, TMI-2 Core Receipt and Storage. The plan described herein is a revision of a previous document entitled Plan for Shipment, Storage, and Examination of TMI-2 Fuel. This revision was required to delineate changes, primarily in Part 2, Core Activities Program, of the previous document. That part of the earlier document related to core examination was reidentified in mid-FY-1984 as a separate trackable entity entitled Core Sample Acquisition and Examination Project, which is not included here

  15. Digital dream analysis: a revised method.

    Science.gov (United States)

    Bulkeley, Kelly

    2014-10-01

    This article demonstrates the use of a digital word search method designed to provide greater accuracy, objectivity, and speed in the study of dreams. A revised template of 40 word search categories, built into the website of the Sleep and Dream Database (SDDb), is applied to four "classic" sets of dreams: The male and female "Norm" dreams of Hall and Van de Castle (1966), the "Engine Man" dreams discussed by Hobson (1988), and the "Barb Sanders Baseline 250" dreams examined by Domhoff (2003). A word search analysis of these original dream reports shows that a digital approach can accurately identify many of the same distinctive patterns of content found by previous investigators using much more laborious and time-consuming methods. The results of this study emphasize the compatibility of word search technologies with traditional approaches to dream content analysis. Copyright © 2014 Elsevier Inc. All rights reserved.

  16. 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

  17. Training Revising Based Traversability Analysis of Complex Terrains for Mobile Robot

    Directory of Open Access Journals (Sweden)

    Rui Song

    2014-05-01

    Full Text Available Traversability analysis is one of the core issues in the autonomous navigation for mobile robots to identify the accessible area by the information of sensors on mobile robots. This paper proposed a model to analyze the traversability of complex terrains based on rough sets and training revising. The model described the traversability for mobile robots by traversability cost. Through the experiment, the paper gets the conclusion that traversability analysis model based on rough sets and training revising can be used where terrain features are rich and complex, can effectively handle the unstructured environment, and can provide reliable and effective decision rules in the autonomous navigation for mobile robots.

  18. 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.

  19. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    CERN Document Server

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  20. Analysis of NRC Regulatory Guide 1.21 Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Yook, Dae Sik; Lee, Byung Soo [KINS, Daejeon (Korea, Republic of)

    2015-05-15

    It is essential to have a degree of uniformity in the methods used for measuring, evaluating, recording, and reporting data on radioactive material in effluents and solid wastes. For this purpose, the U.S. Nuclear Regulatory Commission (NRC) released a revised version of the Regulatory Guide 1.21 'Measuring, evaluating, and reporting radioactive material in liquid and gaseous effluents and solid waste' (revision 2) in 2009, updating the revision 1 version released in 1974. This study compares the previous revision 1 (1974) version with the revision 2 (2009) version to elaborate on the application of the guidelines to Korea. This study consists of an analysis of the 2009 Revision 2 version of the U.S. NRC Regulatory Guidelines 1.21 and an exposition of methods for its application in the domestic environment. Major revisions were made to allow for the adoption of a risk informed approach. Radionuclides with lower than 1% contribution to emission or radiation levels can be selected as principal radionuclides. Requirements for analysis of leaks and spills have been reinforced, with additional groundwater monitoring and hydrological data analysis becoming necessary.

  1. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1985-07-01

    The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

  2. 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

  3. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 2: user's manual (Revision 2)

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1985-07-01

    Revisions to the VIPRE code documents for Volume 2 are presented. These revisions conform to the changes made to VIPRE-01, CYCLE-00 to produce the new version of the code denoted by VIPRE-01, CYCLE-01. The first pages of the revisions specify where the replacement pages are to be inserted and which pages of the original documents should be retained

  4. 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)

  5. Confirmatory Factor Analysis of the Delirium Rating Scale Revised-98 (DRS-R98).

    Science.gov (United States)

    Thurber, Steven; Kishi, Yasuhiro; Trzepacz, Paula T; Franco, Jose G; Meagher, David J; Lee, Yanghyun; Kim, Jeong-Lan; Furlanetto, Leticia M; Negreiros, Daniel; Huang, Ming-Chyi; Chen, Chun-Hsin; Kean, Jacob; Leonard, Maeve

    2015-01-01

    Principal components analysis applied to the Delirium Rating Scale-Revised-98 contributes to understanding the delirium construct. Using a multisite pooled international delirium database, the authors applied confirmatory factor analysis to Delirium Rating Scale-Revised-98 scores from 859 adult patients evaluated by delirium experts (delirium, N=516; nondelirium, N=343). Confirmatory factor analysis found all diagnostic features and core symptoms (cognitive, language, thought process, sleep-wake cycle, motor retardation), except motor agitation, loaded onto factor 1. Motor agitation loaded onto factor 2 with noncore symptoms (delusions, affective lability, and perceptual disturbances). Factor 1 loading supports delirium as a single construct, but when accompanied by psychosis, motor agitation's role may not be solely as a circadian activity indicator.

  6. 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

  7. Revised results for geomechanical testing of MRIG-9 core for the potential SPR siting at the Richton Salt Dome.

    Energy Technology Data Exchange (ETDEWEB)

    Broome, Scott Thomas; Bauer, Stephen J.

    2010-02-01

    This report is a revision of SAND2009-0852. SAND2009-0852 was revised because it was discovered that a gage used in the original testing was mis-calibrated. Following the recalibration, all affected raw data were recalculated and re-presented. Most revised data is similar to, but slightly different than, the original data. Following the data re-analysis, none of the inferences or conclusions about the data or site relative to the SAND2009-0852 data have been changed. A laboratory testing program was developed to examine the mechanical behavior of salt from the Richton salt dome. The resulting information is intended for use in design and evaluation of a proposed Strategic Petroleum Reserve storage facility in that dome. Core obtained from the drill hole MRIG-9 was obtained from the Texas Bureau of Economic Geology. Mechanical properties testing included: (1) acoustic velocity wave measurements; (2) indirect tensile strength tests; (3) unconfined compressive strength tests; (4) ambient temperature quasi-static triaxial compression tests to evaluate dilational stress states at confining pressures of 725, 1450, 2175, and 2900 psi; and (5) confined triaxial creep experiments to evaluate the time-dependent behavior of the salt at axial stress differences of 4000 psi, 3500 psi, 3000 psi, 2175 psi and 2000 psi at 55 C and 4000 psi at 35 C, all at a constant confining pressure of 4000 psi. All comments, inferences, discussions of the Richton characterization and analysis are caveated by the small number of tests. Additional core and testing from a deeper well located at the proposed site is planned. The Richton rock salt is generally inhomogeneous as expressed by the density and velocity measurements with depth. In fact, we treated the salt as two populations, one clean and relatively pure (> 98% halite), the other salt with abundant (at times) anhydrite. The density has been related to the insoluble content. The limited mechanical testing completed has allowed us to

  8. Revised estimates of Greenland ice sheet thinning histories based on ice-core records

    DEFF Research Database (Denmark)

    Lecavalier, B.S.; Milne, G.A.; Fisher, D.A.

    2013-01-01

    -based reconstructions and, to some extent, the estimated elevation histories. A key component of the ice core analysis involved removing the influence of vertical surface motion on the dO signal measured from the Agassiz and Renland ice caps. We re-visit the original analysis with the intent to determine if the use...... of more accurate land uplift curves can account for some of the above noted discrepancy. To improve on the original analysis, we apply a geophysical model of glacial isostatic adjustment calibrated to sea-level records from the Queen Elizabeth Islands and Greenland to calculate the influence of land...... in this selection is further complicated by the possible influence of Innuitian ice during the early Holocene (12-8 ka BP). Our results indicate that a more accurate treatment of the uplift correction leads to elevation histories that are, in general, shifted down relative to the original curves at GRIP, NGRIP, DYE...

  9. Revised and extended analysis of Br IV

    Science.gov (United States)

    Riyaz, A.; Rahimullah, K.; Tauheed, A.

    2014-01-01

    The spectrum of three-times ionized bromine Br IV has been studied in the 319-2350 Å wavelength region. The spectrum was recorded on a 3-m normal incidence vacuum spectrograph at the St. Francis Xavier University, Antigonish (Canada) and 6.65-m grazing incidence spectrograph at the Zeeman laboratory (Amsterdam). The light sources used were a triggered spark and sliding spark, respectively. The ground configuration of Br IV 3d104s24p2, the excited configurations 3d104s4p3+3d104s24p (4d+5d+6d+5s+6s+7s) in the odd parity system and 3d104s24p (5p+4f+5f)+3d104s4p2 (4d+5s)+3d104p4 in the even parity system have been studied. Relativistic Hartree-Fock (HFR) and least squares fitted (LSF) parametric calculations were used to interpret the observed spectrum. 120 Levels of Br IV have now been established, 58 being new. Among 424 spectral lines, 277 are newly classified. The levels 4s4p35S2, 4s24p4d 3F4 and 4p5p (3P0, 1, 3D1, 2, 3S1) are revised. We estimate the accuracy of our measured wavelength for sharp and unblended lines to be ±0.005 Å. The ionization limit is determined as 385,390±100 cm-1 (47.782±0.012 eV).

  10. Meta-Review of CSF core biomarkers in Alzheimer’s disease: the state-of-the-art after the new revised diagnostic criteria

    Directory of Open Access Journals (Sweden)

    Daniel eFerreira

    2014-03-01

    Full Text Available Background: Current research criteria for Alzheimer’s disease (AD include Cerebrospinal Fluid (CSF biomarkers into the diagnostic algorithm. However, spreading their use to the clinical routine is still questionable. Objective: To provide an updated, systematic and critical review on the diagnostic utility of the CSF core biomarkers for AD. Data sources: MEDLINE, PreMedline, EMBASE, PsycInfo, CINAHL, Cochrane Library, and CRD.Eligibility criteria: 1a Systematic reviews with meta-analysis; 1b Primary studies published after the new revised diagnostic criteria; 2 Evaluation of the diagnostic performance of at least one CSF core biomarker.Results: The diagnostic performance of CSF biomarkers is generally satisfactory. They are optimal for discriminating AD patients from healthy controls. Their combination may also be suitable for Mild Cognitive Impairment (MCI prognosis. However, CSF biomarkers fail to distinguish AD from other forms of dementia. Limitations: 1 use of clinical diagnosis as standard instead of pathological postmortem confirmation; 2 variability of methodological aspects; 3 insufficiently long follow-up periods in MCI studies; and 4 lower diagnostic accuracy in primary care compared with memory clinics. Conclusions: Additional work needs to be done to validate the application of CSF core biomarkers as they are proposed in the new revised diagnostic criteria. The use of CSF core biomarkers in clinical routine is more likely if these limitations are overcome. Early diagnosis is going to be of utmost importance when effective pharmacological treatment will be available and the CSF core biomarkers can also be implemented in clinical trials for drug development.

  11. Meta-Review of CSF Core Biomarkers in Alzheimer’s Disease: The State-of-the-Art after the New Revised Diagnostic Criteria

    Science.gov (United States)

    Ferreira, Daniel; Perestelo-Pérez, Lilisbeth; Westman, Eric; Wahlund, Lars-Olof; Sarría, Antonio; Serrano-Aguilar, Pedro

    2014-01-01

    Background: Current research criteria for Alzheimer’s disease (AD) include cerebrospinal fluid (CSF) biomarkers into the diagnostic algorithm. However, spreading their use to the clinical routine is still questionable. Objective: To provide an updated, systematic and critical review on the diagnostic utility of the CSF core biomarkers for AD. Data sources: MEDLINE, PreMedline, EMBASE, PsycInfo, CINAHL, Cochrane Library, and CRD. Eligibility criteria: (1a) Systematic reviews with meta-analysis; (1b) Primary studies published after the new revised diagnostic criteria; (2) Evaluation of the diagnostic performance of at least one CSF core biomarker. Results: The diagnostic performance of CSF biomarkers is generally satisfactory. They are optimal for discriminating AD patients from healthy controls. Their combination may also be suitable for mild cognitive impairment (MCI) prognosis. However, CSF biomarkers fail to distinguish AD from other forms of dementia. Limitations: (1) Use of clinical diagnosis as standard instead of pathological postmortem confirmation; (2) variability of methodological aspects; (3) insufficiently long follow-up periods in MCI studies; and (4) lower diagnostic accuracy in primary care compared with memory clinics. Conclusion: Additional work needs to be done to validate the application of CSF core biomarkers as they are proposed in the new revised diagnostic criteria. The use of CSF core biomarkers in clinical routine is more likely if these limitations are overcome. Early diagnosis is going to be of utmost importance when effective pharmacological treatment will be available and the CSF core biomarkers can also be implemented in clinical trials for drug development. PMID:24715863

  12. Revised and extended analysis of Br IV

    International Nuclear Information System (INIS)

    Riyaz, A.; Rahimullah, K.; Tauheed, A.

    2014-01-01

    The spectrum of three-times ionized bromine Br IV has been studied in the 319–2350 Å wavelength region. The spectrum was recorded on a 3-m normal incidence vacuum spectrograph at the St. Francis Xavier University, Antigonish (Canada) and 6.65-m grazing incidence spectrograph at the Zeeman laboratory (Amsterdam). The light sources used were a triggered spark and sliding spark, respectively. The ground configuration of Br IV 3d 10 4s 2 4p 2 , the excited configurations 3d 10 4s4p 3 +3d 10 4s 2 4p (4d+5d+6d+5s+6s+7s) in the odd parity system and 3d 10 4s 2 4p (5p+4f+5f)+3d 10 4s4p 2 (4d+5s)+3d 10 4p 4 in the even parity system have been studied. Relativistic Hartree–Fock (HFR) and least squares fitted (LSF) parametric calculations were used to interpret the observed spectrum. 120 Levels of Br IV have now been established, 58 being new. Among 424 spectral lines, 277 are newly classified. The levels 4s4p 35 S 2 , 4s 2 4p4d 3 F 4 and 4p5p ( 3 P 0,1 , 3 D 1,2 , 3 S 1 ) are revised. We estimate the accuracy of our measured wavelength for sharp and unblended lines to be ±0.005 Å. The ionization limit is determined as 385,390±100 cm −1 (47.782±0.012 eV). -- Highlights: • The spectrum of Br was recorded on a 3-m spectrograph with triggered spark source. • Atomic transitions for Br IV were identified to established new energy levels. • CI calculations with relativistic corrections were made for theoretical predictions. • Weighted oscillator strength (gf) and transition probabilities (gA) were calculated. • Ionization potential of Br IV was determined experimentally

  13. Preliminary hazards analysis of thermal scrap stabilization system. Revision 1

    International Nuclear Information System (INIS)

    Lewis, W.S.

    1994-01-01

    This preliminary analysis examined the HA-21I glovebox and its supporting systems for potential process hazards. Upon further analysis, the thermal stabilization system has been installed in gloveboxes HC-21A and HC-21C. The use of HC-21C and HC-21A simplified the initial safety analysis. In addition, these gloveboxes were cleaner and required less modification for operation than glovebox HA-21I. While this document refers to glovebox HA-21I for the hazards analysis performed, glovebox HC-21C is sufficiently similar that the following analysis is also valid for HC-21C. This hazards analysis document is being re-released as revision 1 to include the updated flowsheet document (Appendix C) and the updated design basis (Appendix D). The revised Process Flow Schematic has also been included (Appendix E). This Current revision incorporates the recommendations provided from the original hazards analysis as well. The System Design Description (SDD) has also been appended (Appendix H) to document the bases for Safety Classification of thermal stabilization equipment

  14. 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.

  15. Crystallographic Analysis and Structural Revision of a ...

    African Journals Online (AJOL)

    ABSTRACT. Single crystal X-ray analysis of a spiroterpenoid rearrangement product has revealed that its structure is, in fact, isomeric with the structure proposed previously – an observation that has significant mechanistic implications. KEYWORDS. Spiroterpenoid, rearrangement, X-ray crystallography, camphor derivative.

  16. Crystallographic Analysis and Structural Revision of a ...

    African Journals Online (AJOL)

    Single crystal X-ray analysis of a spiroterpenoid rearrangement product has revealed that its structure is, in fact, isomeric with the structure proposed previously – an observation that has significant mechanistic implications. Keywords: Spiroterpenoid, rearrangement, X-ray crystallography, camphor derivative.

  17. Deaf Smith County noise analysis: Revision 2

    International Nuclear Information System (INIS)

    1985-11-01

    An analysis of activities proposed for the three major phases of development of the proposed nuclear waste repository site in Deaf Smith County, Texas, was conducted to quantify the noise levels and the effect of noise resulting from these activities. The report provides additional details of the predictive noise level modeling conducted for the site characterization, repository construction, and repository operation phases. Equivalent day/night sound levels are presented for each phase as sound level contours. Sound levels from onsite and offsite activities are addressed including traffic on access routes, and railroad construction and operation. A description of the predictive models, the analysis methodologies, the noise source inventories, the model outputs, and the evaluation criteria are included. 35 refs., 6 figs., 2 tabs

  18. Richton Dome air quality analysis: Revision 2

    International Nuclear Information System (INIS)

    1985-11-01

    Detailed supporting calculations, methodology and results of the air quality analysis performed for the Richton Dome Environmental Assessment are presented in this report. Maximum emission rates during site characterization and repository construction and operation are analyzed and reported. The major source of emissions is fugitive dust from construction activities. Modeling was performed primarily with the US Environmental Protection Agency Industrial Source Complex (ISC) Model and meteorological data from Jackson, Mississippi. Predicted maximum ground level concentrations off site are presented. Supporting calculations and computer model runs are presented in appendixes. Salt deposition around the site was predicted and results and supporting analyses are presented. 4 refs., 1 fig., 21 tabs

  19. 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.

  20. 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

  1. Final report for Tank 241-B-101, push mode cores 90 and 91. Revision 1

    International Nuclear Information System (INIS)

    Schreiber, R.D.

    1995-11-01

    This is the final report for tank 241-at sign 101, cores 90 and 91. Samples from these cores were analyzed for safety screening purposes in accordance with the Tank 241-B-101 Tank Characterization Plan (TCP) (Reference 1). This final report contains the results for three sets of TGA and gravimetric analyses performed after the 90-day report was issued. Two of these TGA/gravimetric percent water sets of analyses were done because low original TGA results were obtained for the lower half segment of core 90, segment 2 and the facie of core 91, segment 2; the third set of analyses were performed because the TGA and gravimetric percent water results for the upper half segment of core 90, segment 2 differed by approximately a factor of three and further investigation was warranted

  2. 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.

  3. Deep Borehole Emplacement Mode Hazard Analysis Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-07

    This letter report outlines a methodology and provides resource information for the Deep Borehole Emplacement Mode Hazard Analysis (DBEMHA). The main purpose is identify the accident hazards and accident event sequences associated with the two emplacement mode options (wireline or drillstring), to outline a methodology for computing accident probabilities and frequencies, and to point to available databases on the nature and frequency of accidents typically associated with standard borehole drilling and nuclear handling operations. Risk mitigation and prevention measures, which have been incorporated into the two emplacement designs (see Cochran and Hardin 2015), are also discussed. A key intent of this report is to provide background information to brief subject matter experts involved in the Emplacement Mode Design Study. [Note: Revision 0 of this report is concentrated more on the wireline emplacement mode. It is expected that Revision 1 will contain further development of the preliminary fault and event trees for the drill string emplacement mode.

  4. 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)

  5. 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.

  6. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-06-01

    The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

  7. 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)

  8. 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)

  9. 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.

  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. Revised analysis of the Transition Joint Life Test

    Energy Technology Data Exchange (ETDEWEB)

    Sartory, W K

    1984-07-01

    The Transition Joint Life Test was performed by General Electric and the Energy Technology Engineering Center and was analyzed earlier by General Electric. Because of later developments in analysis techniques and in our understanding of the stress behavior near a dissimilar metal weldment, agreement was reached between General Electric and Oak Ridge National Laboratory`s High-Temperature Structural Design program that a more up-to-date analysis was needed. This report presents results of a new analysis incorporating modified mechanical properties, revised constitutive equations, and a more refined finite element grid. The structural life prediction of the present report is quite conservative (approximately 3 cycles of predicted life compared to 12 to 25 cycles of measured life), whereas the earlier General Electric analysis was underconservative by a similar factor.

  12. 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

  13. 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).

  14. 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)

  15. 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)

  16. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

  17. Method for environmental risk analysis (MIRA) revision 2007

    International Nuclear Information System (INIS)

    2007-04-01

    OLF's instruction manual for carrying out environmental risk analyses provides a united approach and a common framework for environmental risk assessments. This is based on the best information available. The manual implies standardizations of a series of parameters, input data and partial analyses that are included in the environmental risk analysis. Environmental risk analyses carried out according to the MIRA method will thus be comparable between fields and between companies. In this revision an update of the text in accordance with today's practice for environmental risk analyses and prevailing regulations is emphasized. Moreover, method adjustments for especially protected beach habitats have been introduced, as well as a general method for estimating environmental risk concerning fish. Emphasis has also been put on improving environmental risk analysis' possibilities to contribute to a better management of environmental risk in the companies (ml)

  18. Phylogenetic analysis and taxonomic revision of Physodactylinae (Coleoptera, Elateridae

    Directory of Open Access Journals (Sweden)

    Simone Policena Rosa

    2014-01-01

    Full Text Available A phylogeny based on male morphological characters and taxonomic revision of the Physodactylinae genera are presented. The phylogenetic analysis based on 66 male characters resulted in the polyphyly of Physodactylinae which comprises four independent lineages. Oligostethius and Idiotropia from Africa were found to be sister groups. Teslasena from Brazil was corroborated as belonging to Cardiophorinae clade. The South American genera Physodactylus and Dactylophysus were found to be sister groups and phylogenetically related to Heterocrepidius species. The Oriental Toxognathus resulted as sister group of that clade plus (Dicrepidius ramicornis (Lissomus sp, Physorhynus erythrocephalus. Taxonomic revisions include diagnoses and redescriptions of genera and distributional records and illustrations of species. Key to species of Teslasena, Toxognathus, Dactylophysus and Physodactylus are also provided. Teslasena lucasi is synonymized with T. femoralis. A new species of Dactylophysus is described, D. hirtus sp. nov., and lectotypes are designated to non-conspecific D. mendax sensu Fleutiaux and Heterocrepidius mendax Candèze. Physodactylus niger is removed from synonymy under P. oberthuri; P. carreti is synonymized with P. niger; P. obesus and P. testaceus are synonymized with P. sulcatus. Nine new species are described in Physodactylus: P. asper sp. nov., P. brunneus sp. nov., P. chassaini sp. nov., P. flavifrons sp. nov., P. girardi sp. nov., P. gounellei sp. nov., P. latithorax sp. nov., P. patens sp. nov. and P. tuberculatus sp. nov.

  19. Revised and extended analysis of trebly ionized bromine: Br IV

    International Nuclear Information System (INIS)

    Tauheed, A; Jabeen, S; Joshi, Y N

    2008-01-01

    The spectrum of Br IV has been studied in the wavelength region 320-1290 A based on the recordings made on a 3-m normal incidence spectrograph at Antigonish Laboratory and on a 6.65-m grazing incidence spectrograph at Zeeman Laboratory, Amsterdam.The light sources used were a triggered spark and sliding spark, respectively. The 4s 2 4p 2 -[4s4p 3 +4s 2 4p (4d+5d+6d+5s+6s+7s)] transition array has been investigated using Hartree-Fock (HF) and least squares fitting (LSF) parametric calculations. Previous analysis has been revised and extended considerably. A value of 1 S 0 level of the ground configuration 4s 2 4p 2 has been found at 33575.5 cm -1 . One hundred and forty-two lines have been classified in this spectrum to establish 61 energy levels out of which 43 are new. The wavelength accuracy of our measurements for sharp lines is ±0.005 A. A revised value of the ionization potential of Br IV has been found to be 385 400±250 cm -1 (47.77±0.03 eV).

  20. An analysis of teacher’s preparation in implementing 2013 revision edition curriculum on mathematics specialization learning

    Science.gov (United States)

    Susilo, T.; Suryawan, A.

    2018-05-01

    This study aimed to determine the pedagogical competence of teachers, the readiness of planning and implementation of learning related to the implementation 2013 revised edition curriculum on mathematics specialization learning for senior high schools Wonogiri. Informants in this study there are 6 high school mathematics teachers X and XI class who teach in the school district Wonogiri. Data were collected using questionnaire method, interview, observation and documentation. Qualitative data analysis is done interactively through 4 paths: data collection, data reduction, data display, drawing conclusion. The results showed that high school mathematics teacher class X and XI in school district of Wonogiri City. The results show that most high school mathematics teachers in grade X and XI are ready to implement the 2013 revised edition curriculum and a few have not been able to implement due to internal or external factors. High school math teachers at Wonogiri district who are ready to face the 2013 revised edition curriculum have applied 10 teacher pedagogic competency indicators according to Regulation of the national education ministry Number 16 Year 2007 in learning. The readiness and implementation of mathematics learning is in line with the demands of the 2013 revised edition curriculum. Based on the teachers who are not ready, data on issues that arise in the implementation of the 2013 revised edition curriculum. Especially the problems in learning, namely mismatch of Core Competence (KI) and Basic Competence (KD) in teacher manual, material disregard in student handbook and lack of examples of problems that exist in teacher manual.

  1. 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.

  2. Spectroscopic properties of reaction center pigments in photosystem II core complexes: revision of the multimer model.

    Science.gov (United States)

    Raszewski, Grzegorz; Diner, Bruce A; Schlodder, Eberhard; Renger, Thomas

    2008-07-01

    Absorbance difference spectra associated with the light-induced formation of functional states in photosystem II core complexes from Thermosynechococcus elongatus and Synechocystis sp. PCC 6803 (e.g., P(+)Pheo(-),P(+)Q(A)(-),(3)P) are described quantitatively in the framework of exciton theory. In addition, effects are analyzed of site-directed mutations of D1-His(198), the axial ligand of the special-pair chlorophyll P(D1), and D1-Thr(179), an amino-acid residue nearest to the accessory chlorophyll Chl(D1), on the spectral properties of the reaction center pigments. Using pigment transition energies (site energies) determined previously from independent experiments on D1-D2-cytb559 complexes, good agreement between calculated and experimental spectra is obtained. The only difference in site energies of the reaction center pigments in D1-D2-cytb559 and photosystem II core complexes concerns Chl(D1). Compared to isolated reaction centers, the site energy of Chl(D1) is red-shifted by 4 nm and less inhomogeneously distributed in core complexes. The site energies cause primary electron transfer at cryogenic temperatures to be initiated by an excited state that is strongly localized on Chl(D1) rather than from a delocalized state as assumed in the previously described multimer model. This result is consistent with earlier experimental data on special-pair mutants and with our previous calculations on D1-D2-cytb559 complexes. The calculations show that at 5 K the lowest excited state of the reaction center is lower by approximately 10 nm than the low-energy exciton state of the two special-pair chlorophylls P(D1) and P(D2) which form an excitonic dimer. The experimental temperature dependence of the wild-type difference spectra can only be understood in this model if temperature-dependent site energies are assumed for Chl(D1) and P(D1), reducing the above energy gap from 10 to 6 nm upon increasing the temperature from 5 to 300 K. At physiological temperature, there are

  3. 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)

  4. Engineering task plan for the annual revision of the rotary mode core sampling system safety equipment list

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This Engineering Task Plan addresses an effort to provide an update to the RMCS Systems 3 and 4 SEL and DCM in order to incorporate the changes to the authorization basis implemented by HNF-SD-WM-BIO-001, Rev. 0 (Draft), Addendum 5 , Safety Analysis for Rotary Mode Core Sampling. Responsibilities, task description, cost estimate, and schedule are presented

  5. 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.)

  6. 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

  7. 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.

  8. Kaplan-Meier Survival Analysis Overestimates the Risk of Revision Arthroplasty: A Meta-analysis.

    Science.gov (United States)

    Lacny, Sarah; Wilson, Todd; Clement, Fiona; Roberts, Derek J; Faris, Peter D; Ghali, William A; Marshall, Deborah A

    2015-11-01

    Although Kaplan-Meier survival analysis is commonly used to estimate the cumulative incidence of revision after joint arthroplasty, it theoretically overestimates the risk of revision in the presence of competing risks (such as death). Because the magnitude of overestimation is not well documented, the potential associated impact on clinical and policy decision-making remains unknown. We performed a meta-analysis to answer the following questions: (1) To what extent does the Kaplan-Meier method overestimate the cumulative incidence of revision after joint replacement compared with alternative competing-risks methods? (2) Is the extent of overestimation influenced by followup time or rate of competing risks? We searched Ovid MEDLINE, EMBASE, BIOSIS Previews, and Web of Science (1946, 1980, 1980, and 1899, respectively, to October 26, 2013) and included article bibliographies for studies comparing estimated cumulative incidence of revision after hip or knee arthroplasty obtained using both Kaplan-Meier and competing-risks methods. We excluded conference abstracts, unpublished studies, or studies using simulated data sets. Two reviewers independently extracted data and evaluated the quality of reporting of the included studies. Among 1160 abstracts identified, six studies were included in our meta-analysis. The principal reason for the steep attrition (1160 to six) was that the initial search was for studies in any clinical area that compared the cumulative incidence estimated using the Kaplan-Meier versus competing-risks methods for any event (not just the cumulative incidence of hip or knee revision); we did this to minimize the likelihood of missing any relevant studies. We calculated risk ratios (RRs) comparing the cumulative incidence estimated using the Kaplan-Meier method with the competing-risks method for each study and used DerSimonian and Laird random effects models to pool these RRs. Heterogeneity was explored using stratified meta-analyses and

  9. Special Analysis: Revision of Saltstone Vault 4 Disposal Limits (U)

    Energy Technology Data Exchange (ETDEWEB)

    Cook, J

    2005-05-26

    New disposal limits have been computed for Vault 4 of the Saltstone Disposal Facility based on several revisions to the models in the existing Performance Assessment and the Special Analysis issued in 2002. The most important changes are the use of a more rigorous groundwater flow and transport model, and consideration of radon emanation. Other revisions include refinement of the aquifer mesh to more accurately model the footprint of the vault, a new plutonium chemistry model accounting for the different transport properties of oxidation states III/IV and V/VI, use of variable infiltration rates to simulate degradation of the closure system, explicit calculation of gaseous releases and consideration of the effects of settlement and seismic activity on the vault structure. The disposal limits have been compared with the projected total inventory expected to be disposed in Vault 4. The resulting sum-of-fractions of the 1000-year disposal limits is 0.2, which indicates that the performance objectives and requirements of DOE 435.1 will not be exceeded. This SA has not altered the conceptual model (i.e., migration of radionuclides from the Saltstone waste form and Vault 4 to the environment via the processes of diffusion and advection) of the Saltstone PA (MMES 1992) nor has it altered the conclusions of the PA (i.e., disposal of the proposed waste in the SDF will meet DOE performance measures). Thus a PA revision is not required and this SA serves to update the disposal limits for Vault 4. In addition, projected doses have been calculated for comparison with the performance objectives laid out in 10 CFR 61. These doses are 0.05 mrem/year to a member of the public and 21.5 mrem/year to an inadvertent intruder in the resident scenario over a 10,000-year time-frame, which demonstrates that the 10 CFR 61 performance objectives will not be exceeded. This SA supplements the Saltstone PA and supersedes the two previous SAs (Cook et al. 2002; Cook and Kaplan 2003).

  10. Core science and technology development plan for indirect-drive ICF ignition. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Powell, H.T.; Kilkenny, J.D. [eds.

    1995-12-01

    To define the development work needed to support inertial confinement fusion (ICF) program goals, the authors have assembled this Core Science and Technology (CS and T) Plan that encompasses nearly all science research and technology development in the ICF program. The objective of the CS and T Plan described here is to identify the development work needed to ensure the success of advanced ICF facilities, in particular the National Ignition Facility (NIF). This plan is intended as a framework to facilitate planning and coordination of future ICF programmatic activities. The CS and T Plan covers all elements of the ICF program including laser technology, optic manufacturing, target chamber, target diagnostics, target design and theory, target components and fabrication, and target physics experiments. The CS and T Plan has been divided into these seven different technology development areas, and they are used as level-1 categories in a work breakdown structure (WBS) to facilitate the organization of all activities in this plan. The scope of the CS and T Plan includes all research and development required to support the NIF leading up to the activation and initial operation as an indirect-drive facility. In each of the CS and T main development areas, the authors describe the technology and issues that need to be addressed to achieve NIF performance goals. To resolve all issues and achieve objectives, an extensive assortment of tasks must be performed in a coordinated and timely manner. The authors describe these activities and present planning schedules that detail the flow of work to be performed over a 10-year period corresponding to estimated time needed to demonstrate fusion ignition with the NIF. Besides the benefits to the ICF program, the authors also discuss how the commercial sector and the nuclear weapons science may profit from the proposed research and development program.

  11. Core science and technology development plan for indirect-drive ICF ignition. Revision 1

    International Nuclear Information System (INIS)

    Powell, H.T.; Kilkenny, J.D.

    1995-12-01

    To define the development work needed to support inertial confinement fusion (ICF) program goals, the authors have assembled this Core Science and Technology (CS and T) Plan that encompasses nearly all science research and technology development in the ICF program. The objective of the CS and T Plan described here is to identify the development work needed to ensure the success of advanced ICF facilities, in particular the National Ignition Facility (NIF). This plan is intended as a framework to facilitate planning and coordination of future ICF programmatic activities. The CS and T Plan covers all elements of the ICF program including laser technology, optic manufacturing, target chamber, target diagnostics, target design and theory, target components and fabrication, and target physics experiments. The CS and T Plan has been divided into these seven different technology development areas, and they are used as level-1 categories in a work breakdown structure (WBS) to facilitate the organization of all activities in this plan. The scope of the CS and T Plan includes all research and development required to support the NIF leading up to the activation and initial operation as an indirect-drive facility. In each of the CS and T main development areas, the authors describe the technology and issues that need to be addressed to achieve NIF performance goals. To resolve all issues and achieve objectives, an extensive assortment of tasks must be performed in a coordinated and timely manner. The authors describe these activities and present planning schedules that detail the flow of work to be performed over a 10-year period corresponding to estimated time needed to demonstrate fusion ignition with the NIF. Besides the benefits to the ICF program, the authors also discuss how the commercial sector and the nuclear weapons science may profit from the proposed research and development program

  12. Revision of failed shoulder hemiarthroplasty to reverse total arthroplasty: analysis of 157 revision implants.

    Science.gov (United States)

    Merolla, Giovanni; Wagner, Eric; Sperling, John W; Paladini, Paolo; Fabbri, Elisabetta; Porcellini, Giuseppe

    2018-01-01

    There remains a paucity of studies examining the conversion of failed hemiarthroplasty (HA) to reverse total shoulder arthroplasty (RTSA). Therefore, the purpose of this study was to examine a large series of revision HA to RTSA. A population of 157 patients who underwent conversion of a failed HA to a revision RTSA from 2006 through 2014 were included. The mean follow-up was 49 months (range, 24-121 months). The indications for revision surgery included instability with rotator cuff insufficiency (n = 127) and glenoid wear (n = 30); instability and glenoid wear were associated in 38 cases. Eight patients with infection underwent 2-stage reimplantation. Patients experienced significant improvements in their preoperative to postoperative pain and shoulder range of motion (P surgeries, secondary to glenoid component loosening (n = 3), instability (n = 3), humeral component disassembly (n = 2), humeral stem loosening (n = 1), and infection (n = 2). Implant survivorship was 95.5% at 2 years and 93.3% at 5 years. There were 4 reoperations including axillary nerve neurolysis (n = 2), heterotopic ossification removal (n = 1), and hardware removal for rupture of the metal cerclage for an acromial fracture (n = 1). At final follow-up, there were 5 "at-risk" glenoid components. Patients experience satisfactory pain relief and recovery of reasonable shoulder function after revision RTSA from a failed HA. There was a relatively low revision rate, with glenoid loosening and instability being the most common causes. Copyright © 2017 Journal of Shoulder and Elbow Surgery Board of Trustees. Published by Elsevier Inc. All rights reserved.

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. Transuranic waste characterization sampling and analysis methods manual. Revision 1

    International Nuclear Information System (INIS)

    Suermann, J.F.

    1996-04-01

    This Methods Manual provides a unified source of information on the sampling and analytical techniques that enable Department of Energy (DOE) facilities to comply with the requirements established in the current revision of the Transuranic Waste Characterization Quality Assurance Program Plan (QAPP) for the Waste Isolation Pilot Plant (WIPP) Transuranic (TRU) Waste Characterization Program (the Program) and the WIPP Waste Analysis Plan. This Methods Manual includes all of the testing, sampling, and analytical methodologies accepted by DOE for use in implementing the Program requirements specified in the QAPP and the WIPP Waste Analysis Plan. The procedures in this Methods Manual are comprehensive and detailed and are designed to provide the necessary guidance for the preparation of site-specific procedures. With some analytical methods, such as Gas Chromatography/Mass Spectrometry, the Methods Manual procedures may be used directly. With other methods, such as nondestructive characterization, the Methods Manual provides guidance rather than a step-by-step procedure. Sites must meet all of the specified quality control requirements of the applicable procedure. Each DOE site must document the details of the procedures it will use and demonstrate the efficacy of such procedures to the Manager, National TRU Program Waste Characterization, during Waste Characterization and Certification audits

  5. Revised analysis of singly ionized xenon, Xe II

    International Nuclear Information System (INIS)

    Hansen, J.E.; Persson, W.

    1987-01-01

    We present a revised analysis of the spectrum of singly ionized xenon, Xe II. This spectrum has been reanalyzed on the basis of the wavelength material published by Drs J. C. Boyce and C. J. Humphreys. The latter has kindly placed the original wavelength list covering the wavelength range 10220-390 A at our disposal. We report 161 energy levels which have been identified on the basis of classifications of 950 lines. We report first f and g levels in Xe II. Also a number of g-factors have been determined for the first time and we give in total 75 g-factors. We have carried out least-squares fits to the even configurations and report the resulting parameter values and eigenvector compositions. A least-squares fit to the 5p 4 6p configuration is also reported. The levels have been named in jK and for many levels also in LS coupling. The former is the better coupling scheme for Xe II. We present an analysis of the 5s photoelectron satellite spectrum of Xe based on our calculated eigenvector compositions and calculations of transition probabilities for ground state transitions as well as lifetimes for the 6p levels. The latter are compared to recent experimental measurements. A list of wavelengths for observed laser transitions showing the present classifications and a discussion of the determination of the ionization potential of Xe II concludes the paper. (orig.)

  6. A reliability analysis of the revised competitiveness index.

    Science.gov (United States)

    Harris, Paul B; Houston, John M

    2010-06-01

    This study examined the reliability of the Revised Competitiveness Index by investigating the test-retest reliability, interitem reliability, and factor structure of the measure based on a sample of 280 undergraduates (200 women, 80 men) ranging in age from 18 to 28 years (M = 20.1, SD = 2.1). The findings indicate that the Revised Competitiveness Index has high test-retest reliability, high inter-item reliability, and a stable factor structure. The results support the assertion that the Revised Competitiveness Index assesses competitiveness as a stable trait rather than a dynamic state.

  7. 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)

  8. 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.

  9. 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.)

  10. 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

  11. Applying revised gap analysis model in measuring hotel service quality.

    Science.gov (United States)

    Lee, Yu-Cheng; Wang, Yu-Che; Chien, Chih-Hung; Wu, Chia-Huei; Lu, Shu-Chiung; Tsai, Sang-Bing; Dong, Weiwei

    2016-01-01

    With the number of tourists coming to Taiwan growing by 10-20 % since 2010, the number has increased due to an increasing number of foreign tourists, particularly after deregulation allowed admitting tourist groups, followed later on by foreign individual tourists, from mainland China. The purpose of this study is to propose a revised gap model to evaluate and improve service quality in Taiwanese hotel industry. Thus, service quality could be clearly measured through gap analysis, which was more effective for offering direction in developing and improving service quality. The HOLSERV instrument was used to identify and analyze service gaps from the perceptions of internal and external customers. The sample for this study included three main categories of respondents: tourists, employees, and managers. The results show that five gaps influenced tourists' evaluations of service quality. In particular, the study revealed that Gap 1 (management perceptions vs. customer expectations) and Gap 9 (service provider perceptions of management perceptions vs. service delivery) were more critical than the others in affecting perceived service quality, making service delivery the main area of improvement. This study contributes toward an evaluation of the service quality of the Taiwanese hotel industry from the perspectives of customers, service providers, and managers, which is considerably valuable for hotel managers. It was the aim of this study to explore all of these together in order to better understand the possible gaps in the hotel industry in Taiwan.

  12. 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).

  13. Regional sensitivity analysis using revised mean and variance ratio functions

    International Nuclear Information System (INIS)

    Wei, Pengfei; Lu, Zhenzhou; Ruan, Wenbin; Song, Jingwen

    2014-01-01

    The variance ratio function, derived from the contribution to sample variance (CSV) plot, is a regional sensitivity index for studying how much the output deviates from the original mean of model output when the distribution range of one input is reduced and to measure the contribution of different distribution ranges of each input to the variance of model output. In this paper, the revised mean and variance ratio functions are developed for quantifying the actual change of the model output mean and variance, respectively, when one reduces the range of one input. The connection between the revised variance ratio function and the original one is derived and discussed. It is shown that compared with the classical variance ratio function, the revised one is more suitable to the evaluation of model output variance due to reduced ranges of model inputs. A Monte Carlo procedure, which needs only a set of samples for implementing it, is developed for efficiently computing the revised mean and variance ratio functions. The revised mean and variance ratio functions are compared with the classical ones by using the Ishigami function. At last, they are applied to a planar 10-bar structure

  14. 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.)

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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)

  20. 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 ...

  1. 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

  2. A Factor Analysis of The Social Interest Index--Revised.

    Science.gov (United States)

    Zarski, John J.; And Others

    1983-01-01

    Factor analyzed the Social Interest Index-Revised (SII-R), which measures levels of social interest attained in each of four life task areas. Four factors (N=308) were defined, i.e., a self-significance factor, a love factor, a friendship factor, and a work factor. Results support the empirical validity of the scale. (Author/PAS)

  3. The Revised Perceived Environmental Control Measure: A Review and Analysis.

    Science.gov (United States)

    Smith-Sebasto, N. J.

    1992-01-01

    A study reveals the need for extensive refinement of the Revised Perceived Environmental Control Measure purported in the past to be a reliable and valid instrument to measure the relationship between the psychological construct, "locus of control," and environmental action or environmentally responsible behavior. (MCO)

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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 ...

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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)

  17. 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)

  18. Comparative Analysis of Direct Hospital Care Costs between Aseptic and Two-Stage Septic Knee Revision

    Science.gov (United States)

    Kasch, Richard; Merk, Sebastian; Assmann, Grit; Lahm, Andreas; Napp, Matthias; Merk, Harry; Flessa, Steffen

    2017-01-01

    Background The most common intermediate and long-term complications of total knee arthroplasty (TKA) include aseptic and septic failure of prosthetic joints. These complications cause suffering, and their management is expensive. In the future the number of revision TKA will increase, which involves a greater financial burden. Little concrete data about direct costs for aseptic and two-stage septic knee revisions with an in depth-analysis of septic explantation and implantation is available. Questions/Purposes A retrospective consecutive analysis of the major partial costs involved in revision TKA for aseptic and septic failure was undertaken to compare 1) demographic and clinical characteristics, and 2) variable direct costs (from a hospital department’s perspective) between patients who underwent single-stage aseptic and two-stage septic revision of TKA in a hospital providing maximum care. We separately analyze the explantation and implantation procedures in septic revision cases and identify the major cost drivers of knee revision operations. Methods A total of 106 consecutive patients (71 aseptic and 35 septic) was included. All direct costs of diagnosis, surgery, and treatment from the hospital department’s perspective were calculated as real purchase prices. Personnel involvement was calculated in units of minutes. Results Aseptic versus septic revisions differed significantly in terms of length of hospital stay (15.2 vs. 39.9 days), number of reported secondary diagnoses (6.3 vs. 9.8) and incision-suture time (108.3 min vs. 193.2 min). The management of septic revision TKA was significantly more expensive than that of aseptic failure ($12,223.79 vs. $6,749.43) (p costs of explantation stage ($4,540.46) were lower than aseptic revision TKA ($6,749.43) which were again lower than those of the septic implantation stage ($7,683.33). All mean costs of stays were not comparable as they differ significantly (p cost drivers were the cost of the implant and

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. Neutron transport and Montecarlo method: analysis and revision

    International Nuclear Information System (INIS)

    Perlado, J.M.

    1982-01-01

    The resolution of the neutron transport equation by the Montecarlo method is presented. Coming from an extensive discussion on the best formulation of that equation in order to be treated through the mentioned method, the theoretical bases of the estimator and random-walk generation is extensively explained. The most general expression for the estimators in different physical situations, each with a diverse random-walk, is included in this basical theoretical part. Furthemore, a large revision on the variance reduction methods is made. Its theoretical presentation is claimed to be in connection with the need for each one of them. The use of the adjoint equation, as a part of the importance sampling, Russian Roulette, splitting, exponential transform, conditional and correlated Montecarlo, and one-collision and next-event extimators, are discussed. Finally, come comments in the presentation of the last works on the theoretical prediction of errors in the generation of estimators-random walks are made. (author)

  6. 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.

  7. 78 FR 45447 - Revisions to Modeling, Data, and Analysis Reliability Standard

    Science.gov (United States)

    2013-07-29

    ...; Order No. 782] Revisions to Modeling, Data, and Analysis Reliability Standard AGENCY: Federal Energy... Analysis (MOD) Reliability Standard MOD- 028-2, submitted to the Commission for approval by the North... Organization. The Commission finds that the proposed Reliability Standard represents an improvement over the...

  8. 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.)

  9. 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.

  10. 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)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. Revision and extension to the analysis of the third spectrum of tellurium: Te III

    International Nuclear Information System (INIS)

    Tauheed, A.; Naz, A.

    2011-01-01

    The spectrum of doubly ionized tellurium atom (Te III) has been investigated in the vacuum ultraviolet wavelength region. The ground configuration of Te III is 5s 2 5p 2 and the excited configurations are of the type 5s 2 5p nl. The core excitation leads to a 5s5p 3 configuration. Cowan's multi-configuration interaction code was utilized to predict the ion structure. The observed spectrum of tellurium was recorded on a 3-m normal incidence vacuum spectrograph of Antigonish Laboratory (Canada) in the wavelength region of 300 - 2000 A by using a triggered spark light source for the excitation of the spectrum. The 5s 2 5p 2 - [ 5s 2 5p (5d + 6d + 7d + 6s + 7s + 8s) + 5s5p 3 ] transition array has been analyzed. Previously reported levels by Joshi et al have been confirmed while the older analysis by Crooker and Joshi has been revised and extended to include the 5s 2 5p (5d, 6d, 7d, 6s,7s, 8s) and 5s5p 3 configurations. Least-squares- fitted parametric calculations were used to interpret the final results. One hundred and fifty spectral lines have been identified to establish 60 energy levels. Our wavelength accuracy for unblended and sharp lines is better than ±0.005 A. The ionization potential of Te III was found to be 224550 ± 300 cm -1 (27.841 ± 0.037eV).

  18. Revised and extended analysis of doubly ionized selenium: Se III

    International Nuclear Information System (INIS)

    Tauheed, A; Hala

    2012-01-01

    The spectrum of selenium was recorded on a 3 m normal incidence vacuum spectrograph of the Antigonish laboratory (Canada) in the wavelength region 300-2080 Å using a triggered spark source. The theoretical structure of doubly ionized selenium (Se III) was predicted by Cowan's multi-configuration interaction code. The ground configuration of Se III is 4s 2 4p 2 and the excited configurations are of the type 4s 2 4pnd (n≥4), 4s 2 4pns (n≥5) and the inner shell excitation gives rise to the 4s4p 3 configuration. The 4s 2 4p 2 -[4s4p 3 +4s 2 4p (4d+5d+6d+7d+5s+6s+7s+8s)] transition array has been analyzed. Several earlier reported levels have been revised and four new configurations have been added. All the levels of these configurations have been established. More than 180 spectral lines have been identified in this spectrum. A total of 75 energy levels belonging to the above-mentioned configurations have been established. Least-squares fitted parametric and Hartree-Fock calculations were used to interpret the observed spectrum. Excellent agreement with theoretical calculations was noticed. The standard deviation of least-squares fit is only 110 cm -1 . The ionization potential of Se III was found to be 255 650±150 cm -1 (31.696±0.018 eV). The accuracy of our wavelengths for sharp lines is better than ±0.005 Å.

  19. Nursing Faculty Decision Making about Best Practices in Test Construction, Item Analysis, and Revision

    Science.gov (United States)

    Killingsworth, Erin Elizabeth

    2013-01-01

    With the widespread use of classroom exams in nursing education there is a great need for research on current practices in nursing education regarding this form of assessment. The purpose of this study was to explore how nursing faculty members make decisions about using best practices in classroom test construction, item analysis, and revision in…

  20. Smart Meters in the Netherlands. Revised financial analysis and policy advice

    International Nuclear Information System (INIS)

    Van Gerwen, R.; Koenis, F.; Schrijner, M.; Widdershoven, G.

    2010-07-01

    Ministry of Economic Affairs, Agriculture and Innovation (ELI) has instructed KEMA to perform a revised cost-benefit analysis to gain insight into the consequences of the changed circumstances with respect to the business case for the introduction of smart meters in the Netherlands.

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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.

  8. 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

  9. Cladistic analysis and revised classification of fossil and recent mysticetes

    DEFF Research Database (Denmark)

    Steeman, Mette Elstrup

    2007-01-01

    A cladistic analysis that includes representatives of all recent genera of mysticetes and several fossil species that were previously referred to the family Cetotheriidae, with tooth-bearing mysticetes and an archaeocete as an outgroup, is presented here. The result of this analysis forms the base...

  10. 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

  11. 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

  12. 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)

  13. 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...

  14. Total hip arthroplasty revision due to infection: a cost analysis approach.

    Science.gov (United States)

    Klouche, S; Sariali, E; Mamoudy, P

    2010-04-01

    The treatment of total hip arthroplasty (THA) infections is long and costly. However,the number of studies in the literature analysing the real cost of THA revision in relation to their etiology, including infection, is limited. The aim of this retrospective study was to determine the cost of revision of infected THA and to compare these costs to those of primary THA and revision of non-infected THA. We performed a retrospective cost analysis for the year 2006 using an identical analytic accounting system in each hospital department (according to internal criteria) based on allotment of direct costs and receipts for each department. From January to December 2006, 424 primary THA, 57 non-infected THA revisions and 40 THA revisions due to infection were performed. The different cost areas of the patient's treatment were identified.This included preoperative medical work-up, medicosurgical management during hospital stay,a second stay in an orthopedic rehabilitation hospital (ORH) and post-hospitalisation antibiotic therapy after revision due to infection, as well as home-based hospitalisation (HH) costs, if this was the selected alternative option. We used the national health insurance fee schedule found in the "Common classification of medical procedures" and the "General nomenclature of professional procedures" applicable in France since September 1, 2005. Hospital costs included direct costs (hospital overhead costs) and indirect costs, (medical, surgical, technical settings and net general service expenses). The calculation of HH costs and ORH costs were based on the average daily charge of these departments. The cost of primary THA was used as the reference.We then compared our surgical costs with those found for the corresponding comparable hospital stay groups (Groupes homogènes de séjour). The average hospital stay (AHS) was 7.5 +/- 1.8 days for primary THA, 8.9 +/- 2.2 days for non-infected revisions and 30.6 +/- 14.9 days for revisions due to infection

  15. Controller Area Network (CAN) schedulability analysis : refuted, revisited and revised

    NARCIS (Netherlands)

    Davis, R.I.; Burns, A.; Bril, R.J.; Lukkien, J.J.

    2007-01-01

    Controller Area Network (CAN) is used extensively in automotive applications, with in excess of 400 million CAN enabled microcontrollers manufactured each year. In 1994 schedulability analysis was developed for CAN, showing how worst-case response times of CAN messages could be calculated and hence

  16. Documentation for the NCES Common Core of Data National Public Education Financial Survey (NPEFS), School Year 2008-09 (Fiscal Year 2009). Revised File Version 1b. NCES 2011-330rev

    Science.gov (United States)

    Cornman, Stephen Q.; Zhou, Lei; Nakamoto, Nanae

    2012-01-01

    This documentation is for the revised file (Version 1b) of the National Center for Education Statistics' (NCES) Common Core of Data (CCD) National Public Education Financial Survey (NPEFS) for school year 2008-2009, fiscal year 2009 (FY 09). It contains a brief description of the data collection along with information required to understand and…

  17. Application of the Intervention Mapping Framework to Develop an Integrated Twenty-First Century Core Curriculum-Part 1: Mobilizing the Community to Revise the Masters of Public Health Core Competencies.

    Science.gov (United States)

    DeBate, Rita; Corvin, Jaime A; Wolfe-Quintero, Kate; Petersen, Donna J

    2017-01-01

    Twenty-first century health challenges have significantly altered the expanding role and functions of public health professionals. Guided by a call from the Association of Schools and Programs of Public Health's (ASPPH) and the Framing the Future: The Second 100 Years of Education for Public Health report to adopt new and innovative approaches to prepare public health leaders, the University of South Florida College of Public Health aimed to self-assess the current Masters of Public Health (MPH) core curriculum with regard to preparing students to meet twenty-first century public health challenges. This paper describes how Intervention Mapping was employed as a framework to increase readiness and mobilize the COPH community for curricular change. Intervention Mapping provides an ideal framework, allowing organizations to access capacity, specify goals, and guide the change process from curriculum development to implementation and evaluation of competency-driven programs. The steps outlined in this paper resulted in a final set of revised MPH core competencies that are interdisciplinary in nature and fulfill the emergent needs to address changing trends in both public health education and challenges in population health approaches. Ultimately, the competencies developed through this process were agreed upon by the entire College of Public Health faculty, signaling one college's readiness for change, while providing the impetus to revolutionize the delivery of public health education at the University of South Florida.

  18. Information Analysis Centers in the Department of Defense. Revision

    Science.gov (United States)

    1987-07-01

    Combat Data Information Center (CDIC) and the Aircraft Survivability Model Repository ( ASMR ) into the Survivability/Vulnerability Information Analysis...Information Center (CDIC) and the Aircraft Survivability Model Respository ( ASMR ). The CDIC was a central repository for combat and test data related to...and ASMR were operated under the technical monitorship of the Flight Dynamics Laboratory at Wright-Patterson AFB, Ohio and were located in Flight

  19. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  20. Dynamic Analysis of Sounding Rocket Pneumatic System Revision

    Science.gov (United States)

    Armen, Jerald

    2010-01-01

    The recent fusion of decades of advancements in mathematical models, numerical algorithms and curve fitting techniques marked the beginning of a new era in the science of simulation. It is becoming indispensable to the study of rockets and aerospace analysis. In pneumatic system, which is the main focus of this paper, particular emphasis will be placed on the efforts of compressible flow in Attitude Control System of sounding rocket.

  1. PUREX Storage Tunnels waste analysis plan. Revision 1

    International Nuclear Information System (INIS)

    Stephenson, M.J.

    1995-11-01

    Washington Administrative Code 173-303-300 requires that a facility develop and follow a written waste analysis plan which describes the procedures that will be followed to ensure that its dangerous waste is managed properly. This document covers the activities at the PUREX Storage Tunnels used to characterize and designate waste that is generated within the PUREX Plant, as well as waste received from other on-site sources

  2. Revised analysis of in-migrating workers during site characterization

    International Nuclear Information System (INIS)

    1987-10-01

    The Deaf Smith Environmental Assessment's analysis of in-migrating workers and community service impacts was predicated on the assumption that a peak of approximately 480 workers would be needed on location to conduct site characterization activities. This analysis assumed that DOE's prime contractor(s) would have a limited staff in the area; the majority of the workers would be on site for the construction of the exploratory shaft and to conduct geologic and environmental studies. Since the time when the Environmental Assessment was prepared, the prime contractors [Battelle-ISSC and the Technical Field Service Contractor (TFSC)] were requested to move their offices to the site area. Therefore, many more administrative and technical workers would be expected to relocate in the Deaf Smith County regions. A change in the expected number of in-migrants could also change the expected nature of community service impacts. It is the purpose of this analysis to evaluate the site characterization workforce and thresholds for local community services. 22 refs., 24 tabs

  3. 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)

  4. A Second-Order Confirmatory Factor Analysis of the Moral Distress Scale-Revised for Nurses.

    Science.gov (United States)

    Sharif Nia, Hamid; Shafipour, Vida; Allen, Kelly-Ann; Heidari, Mohammad Reza; Yazdani-Charati, Jamshid; Zareiyan, Armin

    2017-01-01

    Moral distress is a growing problem for healthcare professionals that may lead to dissatisfaction, resignation, or occupational burnout if left unattended, and nurses experience different levels of this phenomenon. This study aims to investigate the factor structure of the Persian version of the Moral Distress Scale-Revised in intensive care and general nurses. This methodological research was conducted with 771 nurses from eight hospitals in the Mazandaran Province of Iran in 2017. Participants completed the Moral Distress Scale-Revised, data collected, and factor structure assessed using the construct, convergent, and divergent validity methods. The reliability of the scale was assessed using internal consistency (Cronbach's alpha, Theta, and McDonald's omega coefficients) and construct reliability. Ethical considerations: This study was approved by the Ethics Committee of Mazandaran University of Medical Sciences. The exploratory factor analysis ( N = 380) showed that the Moral Distress Scale-Revised has five factors: lack of professional competence at work, ignoring ethical issues and patient conditions, futile care, carrying out the physician's orders without question and unsafe care, and providing care under personal and organizational pressures, which explained 56.62% of the overall variance. The confirmatory factor analysis ( N = 391) supported the five-factor solution and the second-order latent factor model. The first-order model did not show a favorable convergent and divergent validity. Ultimately, the Moral Distress Scale-Revised was found to have a favorable internal consistency and construct reliability. The Moral Distress Scale-Revised was found to be a multidimensional construct. The data obtained confirmed the hypothesis of the factor structure model with a latent second-order variable. Since the convergent and divergent validity of the scale were not confirmed in this study, further assessment is necessary in future studies.

  5. 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

  6. 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.

  7. 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.)

  8. [Quality in Revision Arthroplasty: A Comparison between Claims Data Analysis and External Quality Assurance].

    Science.gov (United States)

    Wessling, M; Gravius, S; Gebert, C; Smektala, R; Günster, C; Hardes, J; Rhomberg, I; Koller, D

    2016-02-01

    External quality assurance for revisions of total knee arthroplasty (TKA) and total hip arthroplasty (THA) are carried out through the AQUA institute in Germany. Data are collected by the providers and are analyzed based on predefined quality indicators from the hospital stay in which the revision was performed. The present study explores the possibility to add routine data analysis to the existing external quality assurance (EQS). Differences between methods are displayed. The study aims to quantify the benefit of an additional analysis that allows patients to be followed up beyond the hospitalization itself. All persons insured in an AOK sickness fund formed the population for analysis. Revisions were identified using the same algorithm as the existing external quality assurance. Adverse events were defined according to the AQUA indicators for the years 2008 to 2011.The hospital stay in which the revision took place and a follow-up of 30 days were included. For re-operation and dislocation we also defined a 365 days interval for additional follow-up. The results were compared to the external quality control reports. Almost all indicators showed higher events in claims data analysis than in external quality control. Major differences are seen for dislocation (EQS SD: 1.87 vs. claims data [cd] SD: 2.06 %, cd+30 d: 2.91 %, cd+365 d: 7.27 %) and reoperation (hip revision: EQS SD: 5.88 % vs. claims data SD: 8.79 % cd+30 d: 9.82 %, cd+365 d: 15.0 %/knee revision: EQS SD: 3.21 % vs. claims data SD: 4.07 %, cd+30 d: 4.6 %, cd+365 d: 15.43 %). Claims data could show additional adverse events for all indicators after the initial hospital stay, rising to 77 % of all events. The number of adverse events differs between the existing external quality control and our claims data analysis. Claims data give the opportunity to complement existing methods of quality control though a longer follow-up, when many complications become evident. Georg

  9. 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)

  10. Replacement Energy Cost Analysis Package (RECAP): User's guide. Revision 1

    International Nuclear Information System (INIS)

    VanKuiken, J.C.; Willing, D.L.

    1994-07-01

    A microcomputer program called the Replacement Energy Cost Analysis Package (RECAP) has been developed to assist the US Nuclear Regulatory Commission (NRC) in determining the replacement energy costs associated with short-term shutdowns or deratings of one or more nuclear reactors. The calculations are based on the seasonal, unit-specific cost estimates for 1993--1996 previously published in NRC Report NUREG/CR--4012, Vol. 3 (1992), for all 112 US reactors. Because the RECAP program is menu-driven, the user can define specific case studies in terms of such parameters as the units to be included, the length and timing of the shutdown or derating period, the unit capacity factors, and the reference year for reporting cost results. In addition to simultaneous shutdown cases, more complicated situations, such as overlapping shutdown periods or shutdowns that occur in different years, can be examined through the use of a present-worth calculation option

  11. Revised Thermal Analysis of LANL Ion Exchange Column

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J

    2006-04-11

    This document updates a previous calculation of the temperature distributions in a Los Alamos National Laboratory (LANL) ion exchange column.1 LANL operates two laboratory-scale anion exchange columns, in series, to extract Pu-238 from nitric acid solutions. The Defense Nuclear Facilities Safety Board has requested an updated analysis to calculate maximum temperatures for higher resin loading capacities obtained with a new formulation of the Reillex HPQ anion exchange resin. The increased resin loading capacity will not exceed 118 g plutonium per L of resin bed. Calculations were requested for normal operation of the resin bed at the minimum allowable solution feed rate of 30 mL/min and after an interruption of flow at the end of the feed stage, when one of the columns is fully loaded. The object of the analysis is to demonstrate that the decay heat from the Pu-238 will not cause resin bed temperatures to increase to a level where the resin significantly degrades. At low temperatures, resin bed temperatures increase primarily due to decay heat. At {approx}70 C a Low Temperature Exotherm (LTE) resulting from the reaction between 8-12 M HNO{sub 3} and the resin has been observed. The LTE has been attributed to an irreversible oxidation of pendant ethyl benzene groups at the termini of the resin polymer chains by nitric acid. The ethyl benzene groups are converted to benzoic acid moities. The resin can be treated to permanently remove the LTE by heating a resin suspension in 8M HNO{sub 3} for 30-45 minutes. No degradation of the resin performance is observed after the LTE removal treatment. In fact, heating the resin in boiling ({approx}115-120 C) 12 M HNO{sub 3} for 3 hr displays thermal stability analogous to resin that has been treated to remove the LTE. The analysis is based on a previous study of the SRS Frames Waste Recovery (FWR) column, performed in support of the Pu-238 production campaign for NASA's Cassini mission. In that study, temperature transients

  12. 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

  13. 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...

  14. 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)

  15. 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

  16. 242-A Evaporator waste analysis plan. Revision 4

    International Nuclear Information System (INIS)

    Basra, T.S.; Mulkey, C.H.

    1994-01-01

    This waste analysis plan (WAP) provides the plan for obtaining information needed for proper waste handling and processing in the 242-A Evaporator located on the Hanford Site. Regulatory and safety issues are addressed by establishing boundary conditions for waste received and treated at the 242-A Evaporator. The boundary conditions are set by establishing limits for items such as potential exothermic reactions, waste compatibility, and control of vessel vent organic emissions. Boundary conditions are also set for operational considerations and to ensure waste acceptance at receiving facilities. The issues that are addressed in this plan include prevention of exotherms in the waste, waste compatibility, vessel vent emissions, and compatibility with the liner in the Liquid Effluent Retention Facility (LERF). The 242-A Evaporator feed stream is separated into two liquid streams: a concentrated slurry stream and a process condensate. A gaseous exhaust stream is also produced. The slurry contains the majority of the radionuclides and inorganic constituents. This stream is pumped back to the double shell tanks (DSTs) and stored for further treatment after being concentrated to target levels. The process condensate (PC) is primarily water that contains trace amounts of organic material and a greatly reduced concentration of radionuclides. The process condensate is presently stored in the (LERF) until it can be further processed in the Effluent Treatment Facility once it is operational

  17. 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

  18. Quality measurement in the shunt treatment of hydrocephalus: analysis and risk adjustment of the Revision Quotient.

    Science.gov (United States)

    Piatt, Joseph H; Freibott, Christina E

    2014-07-01

    OBJECT.: The Revision Quotient (RQ) has been defined as the ratio of the number of CSF shunt revisions to the number of new shunt insertions for a particular neurosurgical practice in a unit of time. The RQ has been proposed as a quality measure in the treatment of childhood hydrocephalus. The authors examined the construct validity of the RQ and explored the feasibility of risk stratification under this metric. The Kids' Inpatient Database for 1997, 2000, 2003, 2006, and 2009 was queried for admissions with diagnostic codes for hydrocephalus and procedural codes for CSF shunt insertion or revision. Revision quotients were calculated for hospitals that performed 12 or more shunt insertions annually. The univariate associations of hospital RQs with a variety of institutional descriptors were analyzed, and a generalized linear model of the RQ was constructed. There were 12,244 admissions (34%) during which new shunts were inserted, and there were 23,349 admissions (66%) for shunt revision. Three hundred thirty-four annual RQs were calculated for 152 different hospitals. Analysis of variance in hospital RQs over the 5 years of study data supports the construct validity of the metric. The following factors were incorporated into a generalized linear model that accounted for 41% of the variance of the measured RQs: degree of pediatric specialization, proportion of initial case mix in the infant age group, and proportion with neoplastic hydrocephalus. The RQ has construct validity. Risk adjustment is feasible, but the risk factors that were identified relate predominantly to patterns of patient flow through the health care system. Possible advantages of an alternative metric, the Surgical Activity Ratio, are discussed.

  19. Frontal sinus revision rate after nasal polyposis surgery including frontal recess clearance and middle turbinectomy: A long-term analysis.

    Science.gov (United States)

    Benkhatar, Hakim; Khettab, Idir; Sultanik, Philippe; Laccourreye, Ollivier; Bonfils, Pierre

    2018-08-01

    To determine the frontal sinus revision rate after nasal polyposis (NP) surgery including frontal recess clearance (FRC) and middle turbinectomy (MT), to search for predictive factors and to analyse surgical management. Longitudinal analysis of 153 patients who consecutively underwent bilateral sphenoethmoidectomy with FRC and MT for NP with a minimum follow-up of 7 years. Decision of revision surgery was made in case of medically refractory chronic frontal sinusitis or frontal mucocele. Univariate and multivariate analysis incorporating clinical and radiological variables were performed. The frontal sinus revision rate was 6.5% (10/153). The mean time between the initial procedure and revision surgery was 3 years, 10 months. Osteitis around the frontal sinus outflow tract (FSOT) was associated with a higher risk of frontal sinus revision surgery (p=0.01). Asthma and aspirin intolerance did not increase the risk, as well as frontal sinus ostium diameter or residual frontoethmoid cells. Among revised patients, 60% required multiple procedures and 70% required frontal sinus ostium enlargement. Our long-term study reports that NP surgery including FRC and MT is associated with a low frontal sinus revision rate (6.5%). Patients developing osteitis around the FSOT have a higher risk of frontal sinus revision surgery. As mucosal damage can lead to osteitis, FSOT mucosa should be preserved during initial NP surgery. However, as multiple procedures are common among NP patients requiring frontal sinus revision, frontal sinus ostium enlargement should be considered during first revision in the hope of reducing the need of further revisions. Copyright © 2018 Elsevier B.V. All rights reserved.

  20. 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.

  1. 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.

  2. 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.

  3. Revision of the ICH guideline on detection of toxicity to reproduction for medicinal products: SWOT analysis.

    Science.gov (United States)

    Barrow, Paul

    2016-09-01

    SWOT analysis was used to gain insights and perspectives into the revision of the ICH S5(R2) guideline on detection of toxicity to reproduction for medicinal products. The current ICH guideline was rapidly adopted worldwide and has an excellent safety record for more than 20 years. The revised guideline should aim to further improve reproductive and developmental (DART) safety testing for new drugs. Alternative methods to animal experiments should be used whenever possible. Modern technology should be used to obtain high quality data from fewer animals. Additions to the guideline should include considerations on the following: limit dose setting, maternal toxicity, biopharmaceuticals, vaccines, testing strategies by indication, developmental immunotoxicity, and male-mediated developmental toxicity. Emerging issues, such as epigenetics and the microbiome, will most likely pose challenges to DART testing in the future. It is hoped that the new guideline will be adopted even outside the ICH regions. Copyright © 2016 Elsevier Inc. All rights reserved.

  4. 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)

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. MODAL ANALYSIS FOR REVISION OF A FEM MODEL OF A STEEL TRUSS BEAM

    Directory of Open Access Journals (Sweden)

    Michal Venglar

    2017-11-01

    Full Text Available The paper deals with a preparation of a complex FEM model for a local damage detection. The initial verified and validated three-dimensional FEM model of a steel truss bridge in laboratory is revised step-by-step to achieve the accurate model according to the experimental model. The emphasis is on modelling of the joints with 4 rivets and modelling of correct boundary conditions, as well as mass parameters and cross-section dimensions.A modal analysis of the structure is performed in FEM software. Many experimental measurements were made to correctly revise the FEM model. The calculated natural frequencies are compared with the measured ones. In addition, mode-shapes from the calculation are validated with the measured mode-shapes. The difference between the prepared FEM model and the measured specimen is small enough after a few steps of tuning. The verified, validated and revised numerical model can be used in future for a local damage detection.

  10. Lawsuits After Primary and Revision Total Hip Arthroplasties: A Malpractice Claims Analysis.

    Science.gov (United States)

    Patterson, Diana C; Grelsamer, Ronald P; Bronson, Michael J; Moucha, Calin S

    2017-10-01

    As the prevalence of total hip arthroplasty (THA) expands, so too will complications and patient dissatisfaction. The goal of this study was to identify the common etiologies of malpractice suits and costs of claims after primary and revision THAs. Analysis of 115 malpractice claims filed for alleged neglectful primary and revision THA surgeries by orthopedic surgeons insured by a large New York state malpractice carrier between 1983 and 2011. The incidence of malpractice claims filed for negligent THA procedures is only 0.15% per year in our population. In primary cases, nerve injury ("foot drop") was the most frequent allegation with 27 claims. Negligent surgery causing dislocation was alleged in 18 and leg length discrepancy in 14. Medical complications were also reported, including 3 thromboembolic events and 6 deaths. In revision cases, dislocation and infection were the most common source of suits. The average indemnity payment was $386,153 and the largest single settlement was $4.1 million for an arterial injury resulting in amputation after a primary hip replacement. The average litigation cost to the insurer was $61,833. Nerve injury, dislocation, and leg length discrepancy are the most common reason for malpractice after primary THA. Orthopedic surgeons should continue to focus on minimizing the occurrence of these complications while adequately incorporating details about the risks and limitations of surgery into their preoperative education. Copyright © 2017 Elsevier Inc. All rights reserved.

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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 ...

  1. 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

  2. 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

  3. A revision of distribution and historical analysis of preferred hosts of Orobanche ramosa (Orobanchaceae in Poland

    Directory of Open Access Journals (Sweden)

    Renata Piwowarczyk

    2012-12-01

    Full Text Available The Polish localities of Orobanche ramosa L., branched broomrape, are either extinct or have not been confirmed for many years. This paper presents two new localities of O. ramosa in Poland from the Płaskowyż Proszowicki plateau (Wyżyna Małopolska upland and the Nizina Nadwiślańska lowland (Kotlina Sandomierska basin. Habitat preferences and the abundance at the sites are described. A revised map of the distribution and a historical analysis of preferred hosts in Poland are included. The taxonomy, biology, ecology and control methods of O. ramosa are also discussed.

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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

  9. 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.

  10. 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)

  11. 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)

  12. 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

  13. Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

    International Nuclear Information System (INIS)

    Yonomoto, T.; Koizumi, Y.; Tasaka, K.

    1987-01-01

    The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

  14. 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)

  15. 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)

  16. 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.)

  17. 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)

  18. 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)

  19. 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)

  20. 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

  1. Predictors of unsuccessful outcome in cemented femoral revisions using bone impaction grafting; Cox regression analysis of 208 cases.

    Science.gov (United States)

    Te Stroet, Martijn A J; Rijnen, Wim H C; Gardeniers, Jean W M; Schreurs, B Willem; Hannink, Gerjon

    2016-09-29

    Despite improvements in the technique of femoral impaction bone grafting, reconstruction failures still can occur. Therefore, the aim of our study was to determine risk factors for the endpoint re-revision for any reason. We used prospectively collected demographic, clinical and surgical data of all 202 patients who underwent 208 femoral revisions using the X-change Femoral Revision System (Stryker-Howmedica), fresh-frozen morcellised allograft and a cemented polished Exeter stem in our department from 1991 to 2007. Univariable and multivariable Cox regression analyses were performed to identify potential factors associated with re-revision. The mean follow-up was 10.6 (5-21) years. The cumulative re-revision rate was 6.3% (13/208). After univariable selection, sex, age, body mass index (BMI), American Association of Anesthesiologists (ASA) classification, type of removed femoral component, and mesh used for reconstruction were included in multivariable regression analysis.In the multivariable analysis, BMI was the only factor that was significantly associated with the risk of re-revision after bone impaction grafting (BMI ≥30 vs. BMI <30, HR = 6.54 [95% CI 1.89-22.65]; p = 0.003). BMI was the only factor associated with the risk of re-revision for any reason. Besides BMI also other factors, such as Endoklinik score and the type of removed femoral component, can provide guidance in the process of preclinical decision making. With the knowledge obtained from this study, preoperative patient selection, informed consent, and treatment protocols can be better adjusted to the individual patient who needs to undergo a femoral revision with impaction bone grafting.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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.)

  8. 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)

  9. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-20

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

  10. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    International Nuclear Information System (INIS)

    1994-01-01

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D ampersand D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities

  11. A Description of the Revised ATHEANA (A Technique for Human Event Analysis)

    International Nuclear Information System (INIS)

    FORESTER, JOHN A.; BLEY, DENNIS C.; COOPER, SUSANE; KOLACZKOWSKI, ALAN M.; THOMPSON, CATHERINE; RAMEY-SMITH, ANN; WREATHALL, JOHN

    2000-01-01

    This paper describes the most recent version of a human reliability analysis (HRA) method called ''A Technique for Human Event Analysis'' (ATHEANA). The new version is documented in NUREG-1624, Rev. 1 [1] and reflects improvements to the method based on comments received from a peer review that was held in 1998 (see [2] for a detailed discussion of the peer review comments) and on the results of an initial trial application of the method conducted at a nuclear power plant in 1997 (see Appendix A in [3]). A summary of the more important recommendations resulting from the peer review and trial application is provided and critical and unique aspects of the revised method are discussed

  12. Method for environmental risk analysis (MIRA) revision 2007; Metode for miljoerettet risikoanalyse (MIRA) revisjon 2007

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    OLF's instruction manual for carrying out environmental risk analyses provides a united approach and a common framework for environmental risk assessments. This is based on the best information available. The manual implies standardizations of a series of parameters, input data and partial analyses that are included in the environmental risk analysis. Environmental risk analyses carried out according to the MIRA method will thus be comparable between fields and between companies. In this revision an update of the text in accordance with today's practice for environmental risk analyses and prevailing regulations is emphasized. Moreover, method adjustments for especially protected beach habitats have been introduced, as well as a general method for estimating environmental risk concerning fish. Emphasis has also been put on improving environmental risk analysis' possibilities to contribute to a better management of environmental risk in the companies (ml)

  13. 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.

  14. Light water reactor fuel analysis code FEMAXI-7. Model and structure [Revised edition

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Amaya, Masaki; Saitou, Hiroaki

    2014-03-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report is the revised edition of the first one which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition, JAEA-Data/Code 2010-035, was published in 2010. The first edition was extended by orderly addition and disposition of explanations of models and organized as the revised edition after three years interval. (author)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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...

  2. 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

  3. 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.

  4. 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)

  5. What Is the Rerevision Rate After Revising a Hip Resurfacing Arthroplasty? Analysis From the AOANJRR.

    Science.gov (United States)

    Wong, James Min-Leong; Liu, Yen-Liang; Graves, Stephen; de Steiger, Richard

    2015-11-01

    More than 15,000 primary hip resurfacing arthroplasties have been recorded by the Australian Orthopaedic Association National Joint Replacement Registry (AOANJRR) with 884 primary procedures requiring revision for reasons other than infection, a cumulative percent revision rate at 12 years of 11%. However, few studies have reported the survivorship of these revision procedures. (1) What is the cumulative percent rerevision rate for revision procedures for failed hip resurfacings? (2) Is there a difference in rerevision rate among different types of revision or bearing surfaces? The AOANJRR collects data on all primary and revision hip joint arthroplasties performed in Australia and after verification against health department data, checking of unmatched procedures, and subsequent retrieval of unreported procedures is able to obtain an almost complete data set relating to hip arthroplasty in Australia. Revision procedures are linked to the known primary hip arthroplasty. There were 15,360 primary resurfacing hip arthroplasties recorded of which 884 had undergone revision and this was the cohort available to study. The types of revisions were acetabular only, femoral only, or revision of both acetabular and femoral components. With the exception of the acetabular-only revisions, all revisions converted hip resurfacing arthroplasties to conventional (stemmed) total hip arthroplasties (THAs). All initial revisions for infection were excluded. The survivorship of the different types of revisions and that of the different bearing surfaces used were estimated using the Kaplan-Meier method and compared using Cox proportional hazard models. Cumulative percent revision was calculated by determining the complement of the Kaplan-Meier survivorship function at that time multiplied by 100. Of the 884 revisions recorded, 102 underwent further revision, a cumulative percent rerevision at 10 years of 26% (95% confidence interval, 19.6-33.5). There was no difference in the rate of

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Revised SRAC code system

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

    1986-09-01

    Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

  12. 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

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. 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...

  18. 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)

  19. 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...

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. Standard Review Plan for the review of safety analysis reports for nuclear power plants, Revision No. 7 to Section 9

    International Nuclear Information System (INIS)

    1978-03-01

    Revision No. 1 to Section 9 of the Standard Review Plan incorporates changes that have been developed since the original issuance in September 1975, many of which are editorial in nature, to reflect current staff practice in the review of safety analysis reports for nuclear power plants

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. [APPLICATION OF COMPUTER-ASSISTED TECHNOLOGY IN ANALYSIS OF REVISION REASON OF UNICOMPARTMENTAL KNEE ARTHROPLASTY].

    Science.gov (United States)

    Jia, Di; Li, Yanlin; Wang, Guoliang; Gao, Huanyu; Yu, Yang

    2016-01-01

    To conclude the revision reason of unicompartmental knee arthroplasty (UKA) using computer-assisted technology so as to provide reference for reducing the revision incidence and improving the level of surgical technique and rehabilitation. The relevant literature on analyzing revision reason of UKA using computer-assisted technology in recent years was extensively reviewed. The revision reasons by computer-assisted technology are fracture of the medial tibial plateau, progressive osteoarthritis of reserved compartment, dislocation of mobile bearing, prosthesis loosening, polyethylene wear, and unexplained persistent pain. Computer-assisted technology can be used to analyze the revision reason of UKA and guide the best operating method and rehabilitation scheme by simulating the operative process and knee joint activities.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. Revised Chapman-Enskog analysis for a class of forcing schemes in the lattice Boltzmann method.

    Science.gov (United States)

    Li, Q; Zhou, P; Yan, H J

    2016-10-01

    In the lattice Boltzmann (LB) method, the forcing scheme, which is used to incorporate an external or internal force into the LB equation, plays an important role. It determines whether the force of the system is correctly implemented in an LB model and affects the numerical accuracy. In this paper we aim to clarify a critical issue about the Chapman-Enskog analysis for a class of forcing schemes in the LB method in which the velocity in the equilibrium density distribution function is given by u=∑_{α}e_{α}f_{α}/ρ, while the actual fluid velocity is defined as u[over ̂]=u+δ_{t}F/(2ρ). It is shown that the usual Chapman-Enskog analysis for this class of forcing schemes should be revised so as to derive the actual macroscopic equations recovered from these forcing schemes. Three forcing schemes belonging to the above class are analyzed, among which Wagner's forcing scheme [A. J. Wagner, Phys. Rev. E 74, 056703 (2006)10.1103/PhysRevE.74.056703] is shown to be capable of reproducing the correct macroscopic equations. The theoretical analyses are examined and demonstrated with two numerical tests, including the simulation of Womersley flow and the modeling of flat and circular interfaces by the pseudopotential multiphase LB model.

  20. Safety analysis on Non-LOCA events for the revision of Wolsong NPP unit 2,3,4 sar

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Ryu, Eui Seung; Kho, Dong Wook; Kim, Sung Min

    2015-01-01

    Korean Wolsong Nuclear Power Plant Units 2,3,4 (CANDU-6 Type) has prepared the revision of safety analysis report (Final Safety Analysis Report (FSAR) chapter 15) from the original performed in the year of 1990s, using the updated and state-of-the-art methodology and tools including IST safety analysis codes and more detail modelling. Compared with the original FSAR15, the revised FSAR15 has significant improvement in both the scope and the depth of safety analysis, which has demonstrated the safety analysis results have complied with the safety requirements(acceptance criteria). This paper will present the analysis scope for Non-LOCA events re-analyzed or added for the FSAR15 revision, methodologies applied such as codes and modelling and some important analysis results will be demonstrated with comparison to acceptance criteria. Application of more detail and near-realistic assumptions and method including Dev-PDO options and uncertainty related to the CHF correlations has altogether brought about more safety margin compared with the original FSAR15 with respect to SDS trip effectiveness etc. (author)

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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)

  7. 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)

  8. 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.

  9. 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)

  10. 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.

  11. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  12. 309 Building fire protection analysis and justification for deactivation of sprinkler system. Revision 1

    International Nuclear Information System (INIS)

    Conner, R.P.

    1997-01-01

    Provide a 'graded approach' fire evaluation in preparation for turnover to Environmental Restoration Contractor for D and D. Scope includes revising 309 Building book value and evaluating fire hazards, radiological and toxicological releases, and life safety issues

  13. Revision and extension to the analysis of the third spectrum of bromine: Br III

    International Nuclear Information System (INIS)

    Jabeen, S.; Tauheed, A.

    2015-01-01

    The spectrum of doubly ionized bromine (Br III) has been investigated in the vacuum ultraviolet wavelength region. Br 2+ is an As-like ion with ground configuration of 4s 2 4p 3 , thus a 3-electron system possessing a complex structure. The theoretical prediction was made using Cowan's quasi-relativistic Hartree–Fock code with superposition of configurations involving the 4s4p 4 , 4s 2 4p 2 (4d+5d+6d+5s+6s+7s), 4s4p 3 (5p+4f), 4p 4 (4d+5s), 4s 2 4p5s5p, 4s4p 2 (4d 2 +5s 2 ), 4s4p 2 4f 2 configurations for the even parity matrix and the 4s 2 4p 3 , 4s 2 4p 2 (5p+6p+4f+5f) configurations for the odd parity matrix. Several previously reported levels of Br III have been revised, and new configurations have been added to the analysis. The spectrum used for this work was recorded on a 3-m normal incidence spectrograph in the wavelength region of 400–1326 Å using a triggered vacuum spark source. One hundred and two energy levels belonging to the 4s 2 4p 3 , 4s4p 4 , 4s 2 4p 2 (4d+5d+6d+5s+6s +7s) configurations have been established, eighty-six being new. Two hundred and seventy-eight lines have been identified in this spectrum. The accuracy of our wavelength measurements for sharp and unblended lines is ±0.006 Å. The ionization potential of Br III was found to be 281,250±100 cm −1 (34.870±0.012 eV). - Highlights: • The spectrum of Br was recorded on a 3-m grating spectrograph with a triggered spark source. • Most of the known energy levels have been revised and further new configurations have been added. • Superposition-of-configurations calculations with relativistic corrections were made for theoretical predictions. • Radiative weighted oscillator strength (gf) & radiative transition probabilities (gA) were calculated. • Ionization Potential of Br III was determined experimentally

  14. 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

  15. 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

  16. 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)

  17. Item Response Theory Analysis of the Psychopathic Personality Inventory-Revised.

    Science.gov (United States)

    Eichenbaum, Alexander E; Marcus, David K; French, Brian F

    2017-06-01

    This study examined item and scale functioning in the Psychopathic Personality Inventory-Revised (PPI-R) using an item response theory analysis. PPI-R protocols from 1,052 college student participants (348 male, 704 female) were analyzed. Analyses were conducted on the 131 self-report items comprising the PPI-R's eight content scales, using a graded response model. Scales collected a majority of their information about respondents possessing higher than average levels of the traits being measured. Each scale contained at least some items that evidenced limited ability to differentiate between respondents with differing levels of the trait being measured. Moreover, 80 items (61.1%) yielded significantly different responses between men and women presumably possessing similar levels of the trait being measured. Item performance was also influenced by the scoring format (directly scored vs. reverse-scored) of the items. Overall, the results suggest that the PPI-R, despite identifying psychopathic personality traits in individuals possessing high levels of those traits, may not identify these traits equally well for men and women, and scores are likely influenced by the scoring format of the individual item and scale.

  18. The Revised Child Anxiety and Depression Scale: A systematic review and reliability generalization meta-analysis.

    Science.gov (United States)

    Piqueras, Jose A; Martín-Vivar, María; Sandin, Bonifacio; San Luis, Concepción; Pineda, David

    2017-08-15

    Anxiety and depression are among the most common mental disorders during childhood and adolescence. Among the instruments for the brief screening assessment of symptoms of anxiety and depression, the Revised Child Anxiety and Depression Scale (RCADS) is one of the more widely used. Previous studies have demonstrated the reliability of the RCADS for different assessment settings and different versions. The aims of this study were to examine the mean reliability of the RCADS and the influence of the moderators on the RCADS reliability. We searched in EBSCO, PsycINFO, Google Scholar, Web of Science, and NCBI databases and other articles manually from lists of references of extracted articles. A total of 146 studies were included in our meta-analysis. The RCADS showed robust internal consistency reliability in different assessment settings, countries, and languages. We only found that reliability of the RCADS was significantly moderated by the version of RCADS. However, these differences in reliability between different versions of the RCADS were slight and can be due to the number of items. We did not examine factor structure, factorial invariance across gender, age, or country, and test-retest reliability of the RCADS. The RCADS is a reliable instrument for cross-cultural use, with the advantage of providing more information with a low number of items in the assessment of both anxiety and depression symptoms in children and adolescents. Copyright © 2017. Published by Elsevier B.V.

  19. Analysis of failed rotator cuff repair – Retrospective survey of revisions after open rotator cuff repair

    Directory of Open Access Journals (Sweden)

    Rupert Schupfner

    2017-07-01

    Full Text Available Background Rotator cuff defects are frequently occurring shoulder pathologies associated with pain and movement impairment. Aims The aim of the study was to analyse the pathologies that lead to operative revisions after primary open rotator cuff repair. Methods In 216 patients who underwent primary rotator cuff repair and later required operative revision between 1996 to 2005, pathologies found intraoperatively during the primary operation and during revision surgery were collected, analysed and compared. Results The average age at the time of revision surgery was 54.3 years. The right shoulder (61.6 per cent was more often affected than the left, males (63.4 per cent more often than females. At primary operation – apart from rotator cuff repair – there were the following surgical procedures performed: 190 acromioplasty, 86 Acromiclavicular joint resections, 68 tenodesis, 40 adhesiolysis and 1 tenotomy. If an ACJ-resection had been performed in the primary operation, ACJ-problems were rare in revision surgery (p<0.01. Primary gleno-humeral adhesions were associated with a significant rise in re-tearing rate (p=0.049. Primary absence of adhesions went along with a significant lower rate of adhesions found at revision (p=0.018. Primary performed acromioplasty had no influence on re-tearing rate (p=0.408 or on the rate of subacromial impingement at revision surgery (p=0.709. Conclusion To avoid operative revision after rotator cuff repair relevant copathologies of the shoulder have to be identified before or during operation and treated accordingly. Therefore, even during open rotator cuff repair, the surgeon should initially start with arthroscopy of the shoulder joint and subacromial space to recognise co-pathologies.

  20. Database tools for enhanced analysis of TMX-U data. Revision 1

    International Nuclear Information System (INIS)

    Stewart, M.E.; Carter, M.R.; Casper, T.A.; Meyer, W.H.; Perkins, D.E.; Whitney, D.M.

    1986-01-01

    A commercial database software package has been used to create several databases and tools that assist and enhance the ability of experimental physicists to analyze data from the Tandem Mirror Experiment-Upgrade (TMX-U) experiment. This software runs on a DEC-20 computer in M-Division's User Service Center at Lawrence Livermore National Laboratory (LLNL), where data can be analyzed offline from the main TMX-U acquisition computers. When combined with interactive data analysis programs, these tools provide the capability to do batch-style processing or interactive data analysis on the computers in the USC or the supercomputers of the National Magnetic Fusion Energy Computer Center (NMFECC) in addition to the normal processing done by the TMX-U acquisition system. One database tool provides highly reduced data for searching and correlation analysis of several diagnostic signals within a single shot or over many shots. A second database tool provides retrieval and storage of unreduced data for use in detailed analysis of one or more diagnostic signals. We will show how these database tools form the core of an evolving offline data analysis environment on the USC computers

  1. 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

  2. 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

  3. 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.)

  4. 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.

  5. 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.

  6. 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.)

  7. 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

  8. 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)

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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...

  14. 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

  15. Beyond factor analysis: Multidimensionality and the Parkinson's Disease Sleep Scale-Revised.

    Directory of Open Access Journals (Sweden)

    Maria E Pushpanathan

    Full Text Available Many studies have sought to describe the relationship between sleep disturbance and cognition in Parkinson's disease (PD. The Parkinson's Disease Sleep Scale (PDSS and its variants (the Parkinson's disease Sleep Scale-Revised; PDSS-R, and the Parkinson's Disease Sleep Scale-2; PDSS-2 quantify a range of symptoms impacting sleep in only 15 items. However, data from these scales may be problematic as included items have considerable conceptual breadth, and there may be overlap in the constructs assessed. Multidimensional measurement models, accounting for the tendency for items to measure multiple constructs, may be useful more accurately to model variance than traditional confirmatory factor analysis. In the present study, we tested the hypothesis that a multidimensional model (a bifactor model is more appropriate than traditional factor analysis for data generated by these types of scales, using data collected using the PDSS-R as an exemplar. 166 participants diagnosed with idiopathic PD participated in this study. Using PDSS-R data, we compared three models: a unidimensional model; a 3-factor model consisting of sub-factors measuring insomnia, motor symptoms and obstructive sleep apnoea (OSA and REM sleep behaviour disorder (RBD symptoms; and, a confirmatory bifactor model with both a general factor and the same three sub-factors. Only the confirmatory bifactor model achieved satisfactory model fit, suggesting that PDSS-R data are multidimensional. There were differential associations between factor scores and patient characteristics, suggesting that some PDSS-R items, but not others, are influenced by mood and personality in addition to sleep symptoms. Multidimensional measurement models may also be a helpful tool in the PDSS and the PDSS-2 scales and may improve the sensitivity of these instruments.

  16. 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.)

  17. Impaction grafting in the femur in cementless modular revision total hip arthroplasty: a descriptive outcome analysis of 243 cases with the MRP-TITAN revision implant

    Directory of Open Access Journals (Sweden)

    Wimmer Matthias D

    2013-01-01

    Full Text Available Abstract Background We present a descriptive and retrospective analysis of revision total hip arthroplasties (THA using the MRP-TITAN stem (Peter Brehm, Weisendorf, GER with distal diaphyseal fixation and metaphyseal defect augmentation. Our hypothesis was that the metaphyseal defect augmentation (Impaction Bone Grafting improves the stem survival. Methods We retrospectively analyzed the aggregated and anonymized data of 243 femoral stem revisions. 68 patients with 70 implants (28.8% received an allograft augmentation for metaphyseal defects; 165 patients with 173 implants (71.2% did not, and served as controls. The mean follow-up was 4.4 ± 1.8 years (range, 2.1–9.6 years. There were no significant differences (p > 0.05 between the study and control group regarding age, body mass index (BMI, femoral defects (types I-III as described by Paprosky, and preoperative Harris Hip Score (HHS. Postoperative clinical function was evaluated using the HHS. Postoperative radiologic examination evaluated implant stability, axial implant migration, signs of implant loosening, periprosthetic radiolucencies, as well as bone regeneration and resorption. Results There were comparable rates of intraoperative and postoperative complications in the study and control groups (p > 0.05. Clinical function, expressed as the increase in the postoperative HHS over the preoperative score, showed significantly greater improvement in the group with Impaction Bone Grafting (35.6 ± 14.3 vs. 30.8 ± 15.8; p ≤ 0.05. The study group showed better outcome especially for larger defects (types II C and III as described by Paprosky and stem diameters ≥ 17 mm. The two groups did not show significant differences in the rate of aseptic loosening (1.4% vs. 2.9% and the rate of revisions (8.6% vs. 11%. The Kaplan-Meier survival for the MRP-TITAN stem in both groups together was 93.8% after 8.8 years. [Study group 95.7% after 8.54 years ; control group 93

  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. A financial analysis of revision hip arthroplasty: the economic burden in relation to the national tariff.

    Science.gov (United States)

    Vanhegan, I S; Malik, A K; Jayakumar, P; Ul Islam, S; Haddad, F S

    2012-05-01

    Revision arthroplasty of the hip is expensive owing to the increased cost of pre-operative investigations, surgical implants and instrumentation, protracted hospital stay and drugs. We compared the costs of performing this surgery for aseptic loosening, dislocation, deep infection and peri-prosthetic fracture. Clinical, demographic and economic data were obtained for 305 consecutive revision total hip replacements in 286 patients performed at a tertiary referral centre between 1999 and 2008. The mean total costs for revision surgery in aseptic cases (n = 194) were £11 897 (sd 4629), for septic revision (n = 76) £21 937 (sd 10 965), for peri-prosthetic fracture (n = 24) £18 185 (sd 9124), and for dislocation (n = 11) £10 893 (sd 5476). Surgery for deep infection and peri-prosthetic fracture was associated with longer operating times, increased blood loss and an increase in complications compared to revisions for aseptic loosening. Total inpatient stay was also significantly longer on average (p < 0.001). Financial costs vary significantly by indication, which is not reflected in current National Health Service tariffs.

  20. 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)

  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. 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)

  11. 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)

  12. 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

  13. 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)

  14. Revision and extension to the analysis of the third spectrum of bromine: Br III

    Science.gov (United States)

    Jabeen, S.; Tauheed, A.

    2015-03-01

    The spectrum of doubly ionized bromine (Br III) has been investigated in the vacuum ultraviolet wavelength region. Br2+ is an As-like ion with ground configuration of 4s24p3, thus a 3-electron system possessing a complex structure. The theoretical prediction was made using Cowan's quasi-relativistic Hartree-Fock code with superposition of configurations involving the 4s4p4, 4s24p2 (4d+5d+6d+5s+6s+7s), 4s4p3 (5p+4f), 4p4(4d+5s), 4s24p5s5p, 4s4p2 (4d2+5s2), 4s4p24f2 configurations for the even parity matrix and the 4s24p3, 4s24p2 (5p+6p+4f+5f) configurations for the odd parity matrix. Several previously reported levels of Br III have been revised, and new configurations have been added to the analysis. The spectrum used for this work was recorded on a 3-m normal incidence spectrograph in the wavelength region of 400-1326 Å using a triggered vacuum spark source. One hundred and two energy levels belonging to the 4s24p3, 4s4p4, 4s24p2 (4d+5d+6d+5s+6s +7s) configurations have been established, eighty-six being new. Two hundred and seventy-eight lines have been identified in this spectrum. The accuracy of our wavelength measurements for sharp and unblended lines is ±0.006 Å. The ionization potential of Br III was found to be 281,250±100 cm-1 (34.870±0.012 eV).

  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. United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    International Nuclear Information System (INIS)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

  18. 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.

  19. 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

  20. 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....

  1. 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…

  2. 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

  3. 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

  4. 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.

  5. 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

  6. Letter of Map Revision

    Data.gov (United States)

    Earth Data Analysis Center, University of New Mexico — The National Flood Hazard Layer (NFHL) data incorporates all Digital Flood Insurance Rate Map(DFIRM) databases published by FEMA, and any Letters Of Map Revision...

  7. 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)

  8. 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...

  9. 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

  10. Detailed analysis of the Japanese version of the Rapid Dementia Screening Test, revised version.

    Science.gov (United States)

    Moriyama, Yasushi; Yoshino, Aihide; Muramatsu, Taro; Mimura, Masaru

    2017-11-01

    The number-transcoding task on the Japanese version of the Rapid Dementia Screening Test (RDST-J) requires mutual conversion between Arabic and Chinese numerals (209 to , 4054 to , to 681, to 2027). In this task, question and answer styles of Chinese numerals are written horizontally. We investigated the impact of changing the task so that Chinese numerals are written vertically. Subjects were 211 patients with very mild to severe Alzheimer's disease and 42 normal controls. Mini-Mental State Examination scores ranged from 26 to 12, and Clinical Dementia Rating scores ranged from 0.5 to 3. Scores of all four subtasks of the transcoding task significantly improved in the revised version compared with the original version. The sensitivity and specificity of total scores ≥9 on the RDST-J original and revised versions for discriminating between controls and subjects with Clinical Dementia Rating scores of 0.5 were 63.8% and 76.6% on the original and 60.1% and 85.8% on revised version. The revised RDST-J total score had low sensitivity and high specificity compared with the original RDST-J for discriminating subjects with Clinical Dementia Rating scores of 0.5 from controls. © 2017 Japanese Psychogeriatric Society.

  11. A Confirmatory Factor Analysis of the Life Orientation Test-Revised with Competitive Athletes

    Science.gov (United States)

    Appaneal, Renee N.

    2012-01-01

    Current reviews outside of sport indicate that the Life Orientation Test-Revised (LOT-R) items load on two separate factors (optimism and pessimism) and, therefore, should be treated as independent constructs. However, researchers in the sport sciences continue to use the single composite score reflecting a unidimensional definition of optimism.…

  12. Peer Scaffolding Behaviors Emerging in Revising a Written Task: A Microgenetic Analysis

    Science.gov (United States)

    Ranjbar, Naser; Ghonsooly, Behzad

    2017-01-01

    Vygotsky's writings on Sociocultural Theory (SCT) of mind, his concept of Zone of Proximal Development (ZPD) and its related metaphor, scaffolding, serve as the theoretical basis for the study of peer collaboration. This paper aimed at examining the effects of peer-scaffolding on EFL writing ability and finding out how revising techniques are…

  13. IGCSE core mathematics

    CERN Document Server

    Wall, Terry

    2013-01-01

    Give your core level students the support and framework they require to get their best grades with this book dedicated to the core level content of the revised syllabus and written specifically to ensure a more appropriate pace. This title has been written for Core content of the revised Cambridge IGCSE Mathematics (0580) syllabus for first teaching from 2013. ? Gives students the practice they require to deepen their understanding through plenty of practice questions. ? Consolidates learning with unique digital resources on the CD, included free with every book. We are working with Cambridge

  14. 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

  15. 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

  16. 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 ...

  17. 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.

  18. Revising Translations

    DEFF Research Database (Denmark)

    Rasmussen, Kirsten Wølch; Schjoldager, Anne

    2011-01-01

    The paper explains the theoretical background and findings of an empirical study of revision policies, using Denmark as a case in point. After an overview of important definitions, types and parameters, the paper explains the methods and data gathered from a questionnaire survey and an interview...... survey. Results clearly show that most translation companies regard both unilingual and comparative revisions as essential components of professional quality assurance. Data indicate that revision is rarely fully comparative, as the preferred procedure seems to be a unilingual revision followed by a more...... or less comparative rereading. Though questionnaire data seem to indicate that translation companies use linguistic correctness and presentation as the only revision parameters, interview data reveal that textual and communicative aspects are also considered. Generally speaking, revision is not carried...

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. 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...

  4. 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

  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 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

  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.

    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

  7. 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.

  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. 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)

  10. 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.

  11. 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.

  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. 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)

  14. 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)

  15. 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)

  16. Analysis on the revision of the United States authorizing procedure for the transfer of unclassified nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Sung-ho; Seo, Hana; Lee, Chansuh; Kim, Jong-sook [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2015-10-15

    The DOE (Department Of Energy) has not comprehensively update 10CFR810 since 1986. Since then, the global civil nuclear market has expanded, particularly in China, the Middle East, and Eastern Europe, with vendors from France, Japan, the Republic of Korea, Russia, and Canada. In result, DOE issued revised 810 in respond to comments received from the public and commercial nuclear market changes. This regulation revision improves the efficiency of authorization process to promote national nuclear industry while maintaining nonproliferation control. Even though ROK has initiated a legal basis for Intangible technology transfer (ITT) for nuclear export control, working implementation system is not set up. This research proposes recommendable ITT implementation of the ROK according to the analysis result of the US regulation. In this revision, of 124 countries had been classified as general authorization under 10CFR810, 80 countries reclassified into the specific authorization. By remaining 'fast track' for specific authorization, in particular, time frames for internal DOE and interagency reviews are reduced. This means the US government actively copes with commercial nuclear market expands to promote their industry. Meanwhile, by remaining some of nuclear-weapon states (China, Russia, India) as specific authorization maintaining that the determinations are consistent with current US national security, diplomatic, and trade policy. By benchmarking the US regulation, Korea can improve the efficiency of the technology transfer authorization process easing the regulatory burden by reducing uncertainty and timelines while maintaining the highest level of nonproliferation control.

  17. Analysis on the revision of the United States authorizing procedure for the transfer of unclassified nuclear technology

    International Nuclear Information System (INIS)

    Yoon, Sung-ho; Seo, Hana; Lee, Chansuh; Kim, Jong-sook

    2015-01-01

    The DOE (Department Of Energy) has not comprehensively update 10CFR810 since 1986. Since then, the global civil nuclear market has expanded, particularly in China, the Middle East, and Eastern Europe, with vendors from France, Japan, the Republic of Korea, Russia, and Canada. In result, DOE issued revised 810 in respond to comments received from the public and commercial nuclear market changes. This regulation revision improves the efficiency of authorization process to promote national nuclear industry while maintaining nonproliferation control. Even though ROK has initiated a legal basis for Intangible technology transfer (ITT) for nuclear export control, working implementation system is not set up. This research proposes recommendable ITT implementation of the ROK according to the analysis result of the US regulation. In this revision, of 124 countries had been classified as general authorization under 10CFR810, 80 countries reclassified into the specific authorization. By remaining 'fast track' for specific authorization, in particular, time frames for internal DOE and interagency reviews are reduced. This means the US government actively copes with commercial nuclear market expands to promote their industry. Meanwhile, by remaining some of nuclear-weapon states (China, Russia, India) as specific authorization maintaining that the determinations are consistent with current US national security, diplomatic, and trade policy. By benchmarking the US regulation, Korea can improve the efficiency of the technology transfer authorization process easing the regulatory burden by reducing uncertainty and timelines while maintaining the highest level of nonproliferation control

  18. 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

  19. 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.

  20. 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.)

  1. 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

  2. Sampling and analysis plan for the gunite and associated tanks interim remedial action, wall coring and scraping at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1998-02-01

    This Sampling and Analysis Plan documents the procedures for collecting and analyzing wall core and wall scraping samples from the Gunite and Associated Tanks. These activities are being conducted to support the Comprehensive Environmental Response, Compensation, and Liability Act at the gunite and associated tanks interim remedial action at Oak Ridge National Laboratory in Oak Ridge, Tennessee. The sampling and analysis activities will be performed in concert with sludge retrieval and sluicing of the tanks. Wall scraping and/or wall core samples will be collected from each quadrant in each tank by using a scraping sampler and/or a coring drill deployed by the Houdini robot vehicle. Each sample will be labeled, transported to the Radioactive Materials Analytical Laboratory, and analyzed for physical and radiological characteristics, including total activity, gross alpha, gross beta, radioactive strontium and cesium, and other alpha- and gamma-emitting radionuclides. The data quality objectives process, based on US Environmental Protection Agency guidance, was applied to identify the objectives of this sampling and analysis. The results of the analysis will be used to (1) validate predictions of a strontium concrete diffusion model, (2) estimate the amount of radioactivity remaining in the tank shells, (3) provide information to correlate with measurements taken by the Gunite Tank Isotope Mapping Probe and the Characterization End Effector, and (4) estimate the performance of the wall cleaning system. This revision eliminates wall-scraping samples from all tanks, except Tank W-3. The Tank W-3 experience indicated that the wall scrapper does not collect sufficient material for analysis

  3. 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

  4. 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

  5. The Morse taper junction in modular revision hip replacement--a biomechanical and retrieval analysis.

    Science.gov (United States)

    Schramm, M; Wirtz, D C; Holzwarth, U; Pitto, R P

    2000-04-01

    All biomaterials used for total joint surgery are subjected to wear mechanisms. Morse taper junctions of modular hip revision implants are predilection sites for both fretting and crevice corrosion, dissociation and breakage of the components. The aim of this study is to quantify wear and study metallurgical changes of Morse taper junctions of in-vitro and in-vivo loaded modular revision stems. Three modular revision stems (MRP-Titan, Peter Brehm GmbH, Germany) were loaded by a servohydraulic testing machine. The loads and conditions used exceeded by far the values required by ISO-standard 7206. The tests were performed with maximum axial loads of 3,500 N to 4,000 N over 10-12 x 10(6) cycles at 2 Hz. Additionally, the female part of the taper junctions were coated with blood and bone debris. The free length of the implant was set to 200 mm. One other MRP stem was investigated after retrieval following 5.5 years of in-vivo use. All contact surfaces of the modular elements were assessed by visual inspection, optical microscopy and scanning electron microscopy (SEM). The degree of plastic deformation of the male part of the morse taper junction was determined by contouroscopy. None of the morse taper junctions broke or failed mechanically. Corrosion and wear affected all tapers, especially at the medial side. The retrieved implant showed no cracks and the amount of debris measured only one third of that for the stems tested in-vitro. The present retrieval and laboratory investigations have proven, that the morse taper junctions of the MRP-titanium stem are stable and resistant to relevant wear mechanisms. The longevity of the junctions for clinical use is given. If an optimal taper design is selected, the advantages of modular femoral components in total hip revision arthroplasty will outweigh the possible risks.

  6. Analysis of the European results on the HTTR's core physics benchmarks

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

    2002-01-01

    Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

  7. 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

  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. 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

  12. 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

  13. 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.

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. Economic analysis of two-stage septic revision after total hip arthroplasty: What are the relevant costs for the hospital's orthopedic department?

    Science.gov (United States)

    Kasch, R; Assmann, G; Merk, S; Barz, T; Melloh, M; Hofer, A; Merk, H; Flessa, S

    2016-03-01

    The number of septic total hip arthroplasty (THA) revisions is increasing continuously, placing a growing financial burden on hospitals. Orthopedic departments performing septic THA revisions have no basis for decision making regarding resource allocation as the costs of this procedure for the departments are unknown. It is widely assumed that septic THA procedures can only be performed at a loss for the department. Therefore, the purpose of this study was to investigate whether this assumption is true by performing a detailed analysis of the costs and revenues for two-stage septic THA revision. Patients who underwent revision THA for septic loosening in two sessions from January 2009 through March 2012 were included in this retrospective, consecutive cost study from the orthopedic department's point of view. We analyzed variable and case-fixed costs for septic revision THA with special regard to implantation and explantation stay. By using marginal costing approach we neglected hospital-fixed costs. Outcome measures include reimbursement and daily contribution margins. The average direct costs (reimbursement) incurred for septic two-stage revision THA was €10,828 (€24,201). The difference in cost and contribution margins per day was significant (p cost for septic revision THA performed in two sessions. Disregarding hospital-fixed costs the included variable and case fixed-costs were covered by revenues. This study provides cost data, which will be guidance for health care decision makers.

  2. A revision of sensitivity analysis for small reactivity effects in ZPRs

    International Nuclear Information System (INIS)

    Ros, Paul; Blaise, Patrick; Gruel, Adrien; Leconte, Pierre

    2017-01-01

    Sensitivity analysis appears to be an important element for nuclear data improvement experiments. Indeed, it brings significant information on the contribution of the isotopes involved in the measurements performed in Zero Power Reactors (ZPRs), particularly oscillation measurements like in MINERVE, and its successor ZEPHYR (Zero power Experimental PHYsics Reactor), currently being designed at CEA. Oscillation measurements consist in oscillating a small sample made of separated isotopes (or irradiated fuels) in the core center. Then, two perturbations occur: a local one corresponding to the flux modification around the sample, and a global one which corresponds to the induced variation of reactivity. This variation of reactivity is either uncontrolled (open loop) or automatically compensated by an external pilot rod (closed loop) to keep the configuration in its critical state. Representativity studies are used in order to evaluate the pertinence of an experiment configuration versus a targeted application. For oscillation experiments, sensitivity of the reactivity effects to nuclear data is needed to obtain such coefficients. The Equivalent Generalized Perturbation Theory (EGPT) method, based on an approximation of the Generalized Perturbation. Theory, is currently applied in the ERANOS code for control rod insertions and other important variations of reactivity. However, such reactivity insertions induce consequent reactivity changes and variations of the flux, whereas oscillations induce maximal reactivity effects of 10 pcm (10 10 -5 Δk/k ) and consequently very local variations of the flux surrounding the sample. Therefore, such numerical methods are not necessarily adapted to the calculation of small reactivity effect sensitivities to nuclear data. The influence of peripheral isotopes (through their cross-sections) to central measurements is evaluated thanks to the deterministic EGPT method and the Monte-Carlo technique of correlated samples. Large

  3. 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

  4. 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

  5. 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...

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. 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)

  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 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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)

  19. Hot sample archiving. Revision 3

    International Nuclear Information System (INIS)

    McVey, C.B.

    1995-01-01

    This Engineering Study revision evaluated the alternatives to provide tank waste characterization analytical samples for a time period as recommended by the Tank Waste Remediation Systems Program. The recommendation of storing 40 ml segment samples for a period of approximately 18 months (6 months past the approval date of the Tank Characterization Report) and then composite the core segment material in 125 ml containers for a period of five years. The study considers storage at 222-S facility. It was determined that the critical storage problem was in the hot cell area. The 40 ml sample container has enough material for approximately 3 times the required amount for a complete laboratory re-analysis. The final result is that 222-S can meet the sample archive storage requirements. During the 100% capture rate the capacity is exceeded in the hot cell area, but quick, inexpensive options are available to meet the requirements

  20. Analysis of the revision of the nuclear cooperation agreement between the USA and Switzerland

    International Nuclear Information System (INIS)

    Lee, K. S.; Lee, B. W.

    1999-01-01

    The US and Switzerland enforced a new nuclear cooperation agreement in June 1998. After extended negotiation for the new nuclear cooperation agreement, Switzerland accomplished to get the advance, long-term consent approach, not to allow the US's prior consent right on the spent fuel used in the US-origin reactors, and to specify conditions for the suspension of advance, long-term consent. This paper analyzes the contents and implications of the revision of the nuclear cooperation agreement between the two countries

  1. 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.

  2. 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.

  3. 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

  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. 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

  7. 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)

  8. 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)

  9. 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

  10. Three or more preoperative injections is the most significant risk factor for revision surgery after operative treatment of lateral epicondylitis: an analysis of 3863 patients.

    Science.gov (United States)

    Degen, Ryan M; Cancienne, Jourdan M; Camp, Christopher L; Altchek, David W; Dines, Joshua S; Werner, Brian C

    2017-04-01

    This study was conducted to identify the rate of failure of operative treatment of lateral epicondylitis, defined as progression to ipsilateral revision surgery, and associated patient-specific risk factors for failure. A national database was used to identify patients undergoing surgical treatment of lateral epicondylitis from 2005 to 2012. Patients undergoing concomitant procedures were excluded. Patients who then required subsequent ipsilateral extensor carpi radialis brevis débridement or release within 2 years were identified using similar methods. A multivariate binomial logistic regression analysis was used to evaluate patient-related risk factors for revision surgery. In addition, the number of preoperative injections (1, 2, or ≥3) in the ipsilateral elbow was identified and included in the regression analysis. Adjusted odds ratios (OR) and 95% confidence intervals were calculated for each risk factor. Of 3863 patients who underwent operative treatment of lateral epicondylitis, 58 (1.5%) required ipsilateral revision surgery. Risk factors for revision surgery included age lateral epicondylitis in the studied population is low (1.5%). Risk factors for revision surgery include younger age, male gender, morbid obesity, tobacco use, and inflammatory arthritis. The most significant risk factor for revision surgery is having ≥3 ipsilateral preoperative injections. Copyright © 2017 Journal of Shoulder and Elbow Surgery Board of Trustees. Published by Elsevier Inc. All rights reserved.

  11. 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

  12. 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

  13. 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)

  14. 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.

  15. 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

  16. 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)

  17. 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)

  18. 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

  19. Representational coexistence in the God concept: Core knowledge intuitions of God as a person are not revised by Christian theology despite lifelong experience.

    Science.gov (United States)

    Barlev, Michael; Mermelstein, Spencer; German, Tamsin C

    2018-01-25

    Previous research has shown that in the minds of young adult religious adherents, acquired theology about the extraordinary characteristics of God (e.g., omniscience) coexists with, rather than replaces, an initial concept of God formed by co-option of the person concept. We tested the hypothesis that representational coexistence holds even after extensive experience with Christian theology, as indexed by age. Christian religious adherents ranging in age from 18 to 87 years were asked to evaluate as true or false statements on which core knowledge intuitions about persons and Christian theology about God were consistent (both true or both false) or inconsistent (true on one and false on the other). Results showed, across adulthood, more theological errors in evaluating inconsistent versus consistent statements. Older adults also exhibited slower response times to inconsistent versus consistent statements. These findings show that despite extensive experience, indeed a lifetime of experience for some participants, the Christian theological God concept does not separate from the initial person concept from which it is formed. In fact, behavioral signatures of representational coexistence were not attenuated by experience. We discuss the broader implications of these findings to the acquisition of evolutionarily new concepts.

  20. An Analysis of Deformities in Revision Surgeries for Secondary Unilateral Cleft Lip

    International Nuclear Information System (INIS)

    Cheema, S. A.; Asim, M.

    2014-01-01

    Objective: To analyze the secondary cleft lip deformities and the possible causes in a cohort of cases. Study Design: A case series. Place and Duration of Study: Services Institute of Medical Sciences and WAPDA Teaching Hospital Complex, Lahore, from September 2008 to March 2012. Methodology: Consecutive cases of secondary unilateral cleft lip deformities were selected for the study. These cases were interviewed and deformities recorded. Pre and postoperative photographs were taken for comparison. Per operative photographs were taken, after marking of the incisions, to keep a record of the intervention needed to correct the deformities. These cases were then further analyzed to know the deformities and interventions needed for correction of these deformities. Results: Study subjects comprised 114 males and 75 females. Secondary correction was the most common in second decade of life with 82 cases in this group. The most common deformity was unfavorable scar in 150 cases followed by notch at the vermilion border in 124 cases. Short lip was found in 119 cases. Complete revision of the repair was required in 158 cases and 25 cases required partial redo of the initial repair. In other 6 cases, only scar revision was carried out. Conclusion: Unfavorable scar followed by vermilion notch and short lip were the most common secondary cleft lip deformities. Better technique helps favorable scar. Vermilion notch and short lip can be overcome by switching from rotation advancement repair to the triangular flap repair of Noordhoff. (author)

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. Analysis of postulated events for the revised ALMR/PRISM design

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.

    1991-01-01

    The PRISM reactor is presently under pre-application licensing review by the NRC, with Brookhaven National Laboratory (BNL) providing technical assistance. In this paper, the authors describe the latest set of results from their SSC and MINET code calculations. Series of postulated events were analyzed, for the most current PRISM design, to evaluate the system performance under unscrammed conditions. The PRISM reactor utilizes a metal fuel composed of 27% Pu, 10% Zr, and 63% U, based on the fuel developed by Argonne National Laboratory as part of the Integral Fast Reactor program. The fuel has a small power and temperature defect as a result of the high fuel thermal conductivity of metal and a hard neutron spectrum. The reactivity feedbacks of the core are designed to provide a negative response to off-normal, high temperature conditions, by invoking negative responses from the thermal expansion of the fuel, control rods, and core radial dimensions. The core restraint system incorporates the limited free bowing feature, which generates an outward bow of the in-core portion of the fuel assembly when a temperature gradient exits. The core also has three Gas Expansion Modules (GEMs) placed around the periphery, which will remove a combined worth of -69 cents of reactivity when the pumps are tripped from full power conditions. In evaluating this passive shutdown response, in which the PRISM reactor power should decrease significantly in response to overheated conditions, one considers three classes of unscrammed events. These include the unscrammed loss of flow (ULOF), loss of heat sink (ULOHS), and transient over power (UTOP) events

  13. 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.

  14. Zero-Headspace Coal-Core Gas Desorption Canister, Revised Desorption Data Analysis Spreadsheets and a Dry Canister Heating System

    Science.gov (United States)

    Barker, Charles E.; Dallegge, Todd A.

    2005-01-01

    Coal desorption techniques typically use the U.S. Bureau of Mines (USBM) canister-desorption method as described by Diamond and Levine (1981), Close and Erwin (1989), Ryan and Dawson (1993), McLennan and others (1994), Mavor and Nelson (1997) and Diamond and Schatzel (1998). However, the coal desorption canister designs historically used with this method have an inherent flaw that allows a significant gas-filled headspace bubble to remain in the canister that later has to be compensated for by correcting the measured desorbed gas volume with a mathematical headspace volume correction (McLennan and others, 1994; Mavor and Nelson, 1997).

  15. 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.

  16. 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

  17. 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

  18. A case of accidental intake of Molybdenum Radionuclides: analysis of data with a revised biokinetic model

    International Nuclear Information System (INIS)

    Giussani, A.; Cantone, M.C.; Tavola, F.; Lopez, M.A.; Navarro, T.

    2002-01-01

    In recent years, a series of investigations on the biokinetics of molybdenum in humans, conducted using stable isotopes of Mo as tracers, has provided valuable experimental data about the dynamics of relevant processes such as the uptake from the gut walls, the clearance from the systemic circulation and the elimination pathways. The results of these studies are in good agreement with the findings of a group of nutritionist who also performed biokinetic studies with stable tracers. All measurements show several deviations from the predictions of the current ICRP model. On the basis of these data, a preliminary revision of the biokinetic model for Mo was presented. The modified model was thus used as a starting point for a new series of biokinetic investigations, aimed at a better definition of some of its features. In a total of 54 studies conducted in 15 volunteers, the influence of the mass and of the chemical form on intestinal absorption, internal kinetics and urinary excretion was investigated

  19. Health effects models for nuclear power plant accident consequence analysis. Part 1, Introduction, integration, and summary: Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S. [Harvard School of Public Health, Boston, MA (United States); Abrahmson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Inhalation Toxicology Research Inst., Albuquerque, NM (United States); Gilbert, E.S. [Battelle Pacific Northwest Lab., Richland, WA (United States)

    1993-10-01

    This report is a revision of NUREG/CR-4214, Rev. 1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. This revision has been made to incorporate changes to the Health Effects Models recommended in two addenda to the NUREG/CR-4214, Rev. 1, Part 11, 1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR, ICRP and NAS/NRC (BEIR V). The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes are considered. Linear and linear-quadratic models are recommended for estimating the risks of seven types of cancer in adults - leukemia, bone, lung, breast, gastrointestinal, thyroid, and ``other``. For most cancers, both incidence and mortality are addressed. Five classes of genetic diseases -- dominant, x-linked, aneuploidy, unbalanced translocations, and multifactorial diseases are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk.

  20. Ocean Optics Protocols for Satellite Ocean Color Sensor Validation, Revision 4, Volume IV: Inherent Optical Properties: Instruments, Characterizations, Field Measurements and Data Analysis Protocols

    Science.gov (United States)

    Mueller, J. L.; Fargion, G. S.; McClain, C. R. (Editor); Pegau, S.; Zanefeld, J. R. V.; Mitchell, B. G.; Kahru, M.; Wieland, J.; Stramska, M.

    2003-01-01

    This document stipulates protocols for measuring bio-optical and radiometric data for the Sensor Intercomparision and Merger for Biological and Interdisciplinary Oceanic Studies (SIMBIOS) Project activities and algorithm development. The document is organized into 6 separate volumes as Ocean Optics Protocols for Satellite Ocean Color Sensor Validation, Revision 4. Volume I: Introduction, Background, and Conventions; Volume II: Instrument Specifications, Characterization and Calibration; Volume III: Radiometric Measurements and Data Analysis Methods; Volume IV: Inherent Optical Properties: Instruments, Characterization, Field Measurements and Data Analysis Protocols; Volume V: Biogeochemical and Bio-Optical Measurements and Data Analysis Methods; Volume VI: Special Topics in Ocean Optics Protocols and Appendices. The earlier version of Ocean Optics Protocols for Satellite Ocean Color Sensor Validation, Revision 3 is entirely superseded by the six volumes of Revision 4 listed above.

  1. Nationwide survey of Arima syndrome: revised diagnostic criteria from epidemiological analysis.

    Science.gov (United States)

    Itoh, Masayuki; Iwasaki, Yuji; Ohno, Kohsaku; Inoue, Takehiko; Hayashi, Masaharu; Ito, Shuichi; Matsuzaka, Tetsuo; Ide, Shuhei; Arima, Masataka

    2014-05-01

    We have never known any epidemiological study of Arima syndrome since it was first described in 1971. To investigate the number of Arima syndrome patients and clarify the clinical differences between Arima syndrome and Joubert syndrome, we performed the first nationwide survey of Arima syndrome, and herein report its results. Furthermore, we revised the diagnostic criteria for Arima syndrome. As a primary survey, we sent out self-administered questionnaires to most of the Japanese hospitals with a pediatric clinic, and facilities for persons with severe motor and intellectual disabilities, inquiring as to the number of patients having symptoms of Arima syndrome, including severe psychomotor delay, agenesis or hypoplasia of cerebellar vermis, renal dysfunction, visual dysfunction and with or without ptosis-like appearance. Next, as the second survey, we sent out detailed clinical questionnaires to the institutes having patients with two or more typical symptoms. The response rate of the primary survey was 72.7% of hospitals with pediatric clinic, 63.5% of national hospitals and 66.7% of municipal and private facilities. The number of patients with 5 typical symptoms was 13 and that with 2-4 symptoms was 32. The response rate of the secondary survey was 52% (23 patients). After reviewing clinical features of 23 patients, we identified 7 Arima syndrome patients and 16 Joubert syndrome patients. Progressive renal dysfunction was noticed in all Arima syndrome patients, but in 33% of those with Joubert syndrome. It is sometimes difficult to distinguish Arima syndrome from Joubert syndrome. Some clinicians described a patient with Joubert syndrome and its complications of visual dysfunction and renal dysfunction, whose current diagnosis was Arima syndrome. Thus, the diagnosis of the two syndromes may be confused. Here, we revised the diagnostic criteria for Arima syndrome. Copyright © 2013 The Japanese Society of Child Neurology. Published by Elsevier B.V. All rights

  2. 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

  3. 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)

  4. 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

  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. 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...

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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 condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

  17. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  18. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    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 energy 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 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  19. Analysis of hypothetical LMFBR whole-core accidents in the USA

    International Nuclear Information System (INIS)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

  20. Global shielding analysis of the 2-element ANS core and reflector with photoneutrons

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    1996-01-01

    This paper describes the initial global 2-D shielding analyses for the 2-element, heavy-water cooled and reflected Advanced Neutron Source reactor which was to have been built in Oak Ridge, Tennessee. The portion of the system analyzed encompassed the highly enriched core, the 1.5-m-thick heavy-water reflector, the aluminum reflector vessel, and the first 0.2 m of light water beyond the reflector vessel. While some results are presented, this paper focuses primarily on the lessons learned during the analysis of this rather unique system

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  3. Development of a detailed core flow analysis code for prismatic fuel reactors

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1990-01-01

    The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

  4. Trochanteric entry femoral nails yield better femoral version and lower revision rates-A large cohort multivariate regression analysis.

    Science.gov (United States)

    Yoon, Richard S; Gage, Mark J; Galos, David K; Donegan, Derek J; Liporace, Frank A

    2017-06-01

    Intramedullary nailing (IMN) has become the standard of care for the treatment of most femoral shaft fractures. Different IMN options include trochanteric and piriformis entry as well as retrograde nails, which may result in varying degrees of femoral rotation. The objective of this study was to analyze postoperative femoral version between three types of nails and to delineate any significant differences in femoral version (DFV) and revision rates. Over a 10-year period, 417 patients underwent IMN of a diaphyseal femur fracture (AO/OTA 32A-C). Of these patients, 316 met inclusion criteria and obtained postoperative computed tomography (CT) scanograms to calculate femoral version and were thus included in the study. In this study, our main outcome measure was the difference in femoral version (DFV) between the uninjured limb and the injured limb. The effect of the following variables on DFV and revision rates were determined via univariate, multivariate, and ordinal regression analyses: gender, age, BMI, ethnicity, mechanism of injury, operative side, open fracture, and table type/position. Statistical significance was set at pregression analysis revealed that a lower BMI was significantly associated with a lower DFV (p=0.006). Controlling for possible covariables, multivariate analysis yielded a significantly lower DFV for trochanteric entry nails than piriformis or retrograde nails (7.9±6.10 vs. 9.5±7.4 vs. 9.4±7.8°, pregression analysis. However, this is not to state that the other nail types exhibited abnormal DFV. Translation to the clinical impact of a few degrees of DFV is also unknown. Future studies to more in-depth study the intricacies of femoral version may lead to improved technology in addition to potentially improved clinical outcomes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. 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

  6. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  7. Control rod drop accident analysis for the mixed core project in Ling Ao NPS

    International Nuclear Information System (INIS)

    Zhang Shishun; Zhou Zhou; Xiao Min

    2004-01-01

    AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

  8. Development of a detailed core flow analysis code for prismatic fuel reactors

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1990-01-01

    The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

  9. Simplified inelastic analysis methods applied to fast breeder reactor core design

    International Nuclear Information System (INIS)

    Abo-El-Ata, M.M.

    1978-01-01

    The paper starts with a review of some currently available simplified inelastic analysis methods used in elevated temperature design for evaluating plastic and thermal creep strains. The primary purpose of the paper is to investigate how these simplified methods may be applied to fast breeder reactor core design where neutron irradiation effects are significant. One of the problems discussed is irradiation-induced creep and its effect on shakedown, ratcheting, and plastic cycling. Another problem is the development of swelling-induced stress which is an additional loading mechanism and must be taken into account. In this respect an expression for swelling-induced stress in the presence of irradiation creep is derived and a model for simplifying the stress analysis under these conditions is proposed. As an example, the effects of irradiation creep and swelling induced stress on the analysis of a thin walled tube under constant internal pressure and intermittent heat fluxes, simulating a fuel pin, is presented

  10. Optimal design for crosstalk analysis in 12-core 5-LP mode homogeneous multicore fiber for different lattice structure

    Science.gov (United States)

    Kumar, Dablu; Ranjan, Rakesh

    2018-03-01

    12-Core 5-LP mode homogeneous multicore fibers have been proposed for analysis of inter-core crosstalk and dispersion, with four different lattice structures (circular, 2-ring, square lattice, and triangular lattice) having cladding diameter of 200 μm and a fixed cladding thickness of 35 μm. The core-to-core crosstalk impact has been studied numerically with respect to bending radius, core pitch, transmission distance, wavelength, and core diameter for all 5-LP modes. In anticipation of further reduction in crosstalk levels, the trench-assisted cores have been incorporated for all respective designs. Ultra-low crosstalk (-138 dB/100 km) has been achieved through the triangular lattice arrangement, with trench depth Δ2 = -1.40% for fundamental (LP01) mode. It has been noted that the impact of mode polarization on crosstalk behavior is minor, with difference in crosstalk levels between two polarized spatial modes as ≤0.2 dB. Moreover, the optimized cladding diameter has been obtained for all 5-LP modes for a target value of crosstalk of -50 dB/100 km, with all the core arrangements. The dispersion characteristic has also been analyzed with respect to wavelength, which is nearly 2.5 ps/nm km at operating wavelength 1550 nm. The relative core multiplicity factor (RCMF) for the proposed design is obtained as 64.

  11. Feasibility analysis of real-time physical modeling using WaveCore processor technology on FPGA

    NARCIS (Netherlands)

    Verstraelen, Martinus Johannes Wilhelmina; Pfeifle, Florian; Bader, Rolf

    2015-01-01

    WaveCore is a scalable many-core processor technology. This technology is specifically developed and optimized for real-time acoustical modeling applications. The programmable WaveCore soft-core processor is silicon-technology independent and hence can be targeted to ASIC or FPGA technologies. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

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

  13. Nondestructive X-Ray Computed Tomography Analysis of Sediment Cores: A Case Study from the Arctic Ocean

    Science.gov (United States)

    Oti, E.; Polyak, L. V.; Cook, A.; Dipre, G.

    2014-12-01

    Investigation of marine sediment records can help elucidate recent changes in the Arctic Ocean circulation and sea ice conditions. We examine sediment cores from the western Arctic Ocean, representing Late to Early Quaternary age (potentially up to 1 Ma). Previous studies of Arctic sediment cores indicate that interglacial/interstadial periods with relatively high sea levels and reduced ice cover are characterized by vigorous bioturbation, while glacial intervals have little to no bioturbation. Traditional methods for studying bioturbation require physical dissection of the cores, effectively destroying them. To treat this limitation, we evaluate archival sections of the cores using an X-ray Computed Tomography (XCT) scanner, which noninvasively images the sediment cores in three dimensions. The scanner produces density sensitive images suitable for quantitative analysis and for identification of bioturbation based on size, shape, and orientation. We use image processing software to isolate burrows from surrounding sediment, reconstruct them three-dimensionally, and then calculate their surface areas, volumes, and densities. Preliminary analysis of a core extending to the early Quaternary shows that bioturbation ranges from 0 to approximately 20% of the core's volume. In future research, we will quantitatively define the relationship between bioturbation activity and glacial regimes. XCT examination of bioturbation and other sedimentary features has the potential to shed light on paleoceanographic conditions such as sedimentation patterns and food flux. XCT is an alternative, underexplored investigation method that bears implications not only for illustrating paleoclimate variations but also for preserving cores for future, more advanced technologies.

  14. MEGA-CC: computing core of molecular evolutionary genetics analysis program for automated and iterative data analysis.

    Science.gov (United States)

    Kumar, Sudhir; Stecher, Glen; Peterson, Daniel; Tamura, Koichiro

    2012-10-15

    There is a growing need in the research community to apply the molecular evolutionary genetics analysis (MEGA) software tool for batch processing a large number of datasets and to integrate it into analysis workflows. Therefore, we now make available the computing core of the MEGA software as a stand-alone executable (MEGA-CC), along with an analysis prototyper (MEGA-Proto). MEGA-CC provides users with access to all the computational analyses available through MEGA's graphical user interface version. This includes methods for multiple sequence alignment, substitution model selection, evolutionary distance estimation, phylogeny inference, substitution rate and pattern estimation, tests of natural selection and ancestral sequence inference. Additionally, we have upgraded the source code for phylogenetic analysis using the maximum likelihood methods for parallel execution on multiple processors and cores. Here, we describe MEGA-CC and outline the steps for using MEGA-CC in tandem with MEGA-Proto for iterative and automated data analysis. http://www.megasoftware.net/.

  15. A seismic analysis of Korean standard PWR fuels under transition core conditions

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

    2005-01-01

    The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

  16. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  17. Analysis of postulated events for the revised ALMR/PRISM design

    International Nuclear Information System (INIS)

    Slovik, G.C.; Tuyle, G.J. Van

    1992-01-01

    The Nuclear Regulatory Commission (NRC) is continuing a pre-application review of the 471 MWt, liquid metal reactor, PRISM, by General Electric, with Brookhaven National Laboratory providing technical support. The revised design has been evaluated, using the SSC code, for an unscrammed loss of heat sink (ULOHS), an unscrammed loss of flow (ULOF) with and without the Gas Expansion Modules (GEMs), and a 40 cents unscrammed transient overpower (UTOP) event. The feedback effects for U-27Pu-10Zr metal fuel were modeled in SSC. The ULOHS accident was determined to be a benign event for the design, with the reactor power transitioning down to a decay heat level within 500s. The power during the postulated ULOF events, with the GEMs functioning, transitioned to decay heat levels without fuel damage, and included a 300K margin to sodium saturation. The case without the GEMs had only a 160K margin to sodium saturation and higher fuel temperatures. In addition, the clad was predicted to quickly pass through the eutectic phase (between fuel and clad), and some clad wastage would result. The 40 cents UTOP was predicted to raise the power to 1.8 rated, which later stabilized near 1.2 times full power. SSC predicted some localized fuel melting for the event, but the significance of this localized damage has not yet been determined. If necessary, the vendor has options to reduce the maximum reactivity insertion below 40 cents

  18. Analysis of postulated events for the revised ALMR/PRISM design

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.

    1991-01-01

    The Nuclear Regulatory Commission (NRC) is continuing a pre-application review of the 471 MWt, Liquid Metal Reactor, PRISM by General Electric, with Brookhaven National Laboratory providing technical support. The revised design has been evaluated, using the SSC code, for an unscrammed loss of heat sink (ULOHS), an unscrammed loss of flow (ULOF) with and without the Gas Expansion Modules (GEMs), and a 40 cents unscrammed transient overpower (UTOP) event. The feedback effects for U-27Pu-10Zr metal fuel were modeled in SSC. The ULOHS accident was determined to be a benign event for the design, with the reactor power transitioning down to a decay heat level within 500s. The power during the postulated ULOF events, with the GEMs functioning, transitioned to decay heat levels without fuel damage, and included a 300K margin to sodium saturation. The case without the GEMs had only a 160K margin to sodium saturation and higher fuel temperatures. In addition, the clad was predicted to quickly pass through the eutectic phase (between fuel and clad), and some clad wastage would result. The 40 cents UTOP was predicted to raise the power to 1.8 rated, which later stabilized near 1.2 times full power. SSC predicted some localized fuel melting for the event, but the significance of this localized damage has not yet been determined. If necessary, the vendor has options to reduce the maximum reactivity insertion below 40 cents

  19. Revised and extended analysis of the fifth spectrum of bromine: Br V

    International Nuclear Information System (INIS)

    Tauheed, A; Joshi, Y N

    2009-01-01

    The spectrum of four-times ionized bromine (Br V) has been investigated in the 350-1850 A wavelength region, which was recorded on a 3-m normal incidence spectrograph at the Antigonish laboratory using a triggered spark light source and on a 6.65-m grazing incidence spectrograph at the Zeeman laboratory in Amsterdam using a sliding spark source. We have studied the 4s 2 (4p+5p+6p+4f)+4p 3 +4s4p4d+4s4p5s configurations in the odd parity system and the 4s4p 2 , 4s 2 (4d+5d+5s+6s+7s) configurations in the even parity system. Relativistic Hartree-Fock (HFR) and least squares fitted (LSF) parametric calculations were carried out to interpret the observed spectrum. Fifty-eight Br V levels have now been established, out of which 48 are new levels. Two hundred and sixteen spectral lines have been classified in this spectrum, out of which 202 are newly classified lines. A revised value of the ionization potential has been determined to be 480 670±200 cm -1 (59.60±0.02 eV).

  20. Revised Human Health Risk Assessment on Chlorpyrifos

    Science.gov (United States)

    We have revised our human health risk assessment and drinking water exposure assessment for chlorpyrifos that supported our October 2015 proposal to revoke all food residue tolerances for chlorpyrifos. Learn about the revised analysis.