Identifying ELIXIR Core Data Resources [version 1; referees: 2 approved

Directory of Open Access Journals (Sweden)

Christine Durinx

2016-09-01

Full Text Available The core mission of ELIXIR is to build a stable and sustainable infrastructure for biological information across Europe. At the heart of this are the data resources, tools and services that ELIXIR offers to the life-sciences community, providing stable and sustainable access to biological data. ELIXIR aims to ensure that these resources are available long-term and that the life-cycles of these resources are managed such that they support the scientific needs of the life-sciences, including biological research. ELIXIR Core Data Resources are defined as a set of European data resources that are of fundamental importance to the wider life-science community and the long-term preservation of biological data. They are complete collections of generic value to life-science, are considered an authority in their field with respect to one or more characteristics, and show high levels of scientific quality and service. Thus, ELIXIR Core Data Resources are of wide applicability and usage. This paper describes the structures, governance and processes that support the identification and evaluation of ELIXIR Core Data Resources. It identifies key indicators which reflect the essence of the definition of an ELIXIR Core Data Resource and support the promotion of excellence in resource development and operation. It describes the specific indicators in more detail and explains their application within ELIXIR’s sustainability strategy and science policy actions, and in capacity building, life-cycle management and technical actions. Establishing the portfolio of ELIXIR Core Data Resources and ELIXIR Services is a key priority for ELIXIR and publicly marks the transition towards a cohesive infrastructure.

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)

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)

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.

Identifying ELIXIR Core Data Resources [version 2; referees: 2 approved

Directory of Open Access Journals (Sweden)

Christine Durinx

2017-03-01

Full Text Available The core mission of ELIXIR is to build a stable and sustainable infrastructure for biological information across Europe. At the heart of this are the data resources, tools and services that ELIXIR offers to the life-sciences community, providing stable and sustainable access to biological data. ELIXIR aims to ensure that these resources are available long-term and that the life-cycles of these resources are managed such that they support the scientific needs of the life-sciences, including biological research. ELIXIR Core Data Resources are defined as a set of European data resources that are of fundamental importance to the wider life-science community and the long-term preservation of biological data. They are complete collections of generic value to life-science, are considered an authority in their field with respect to one or more characteristics, and show high levels of scientific quality and service. Thus, ELIXIR Core Data Resources are of wide applicability and usage. This paper describes the structures, governance and processes that support the identification and evaluation of ELIXIR Core Data Resources. It identifies key indicators which reflect the essence of the definition of an ELIXIR Core Data Resource and support the promotion of excellence in resource development and operation. It describes the specific indicators in more detail and explains their application within ELIXIR’s sustainability strategy and science policy actions, and in capacity building, life-cycle management and technical actions. The identification process is currently being implemented and tested for the first time. The findings and outcome will be evaluated by the ELIXIR Scientific Advisory Board in March 2017. Establishing the portfolio of ELIXIR Core Data Resources and ELIXIR Services is a key priority for ELIXIR and publicly marks the transition towards a cohesive infrastructure.

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.

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)

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)

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.

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)

Identifying Core Vocabulary for Urdu Language Speakers Using Augmentative Alternative Communication

Science.gov (United States)

Mukati, Abdul Samad

2013-01-01

The purpose of this research is to identify a core set of vocabulary used by native Urdu language (UL) speakers during dyadic conversation for social interaction and relationship building. This study was conducted in Karachi, Pakistan at an institution of higher education. This research seeks to distinguish between general (nonspecific…

Identifying protein complex by integrating characteristic of core-attachment into dynamic PPI network.

Directory of Open Access Journals (Sweden)

Xianjun Shen

Full Text Available How to identify protein complex is an important and challenging task in proteomics. It would make great contribution to our knowledge of molecular mechanism in cell life activities. However, the inherent organization and dynamic characteristic of cell system have rarely been incorporated into the existing algorithms for detecting protein complexes because of the limitation of protein-protein interaction (PPI data produced by high throughput techniques. The availability of time course gene expression profile enables us to uncover the dynamics of molecular networks and improve the detection of protein complexes. In order to achieve this goal, this paper proposes a novel algorithm DCA (Dynamic Core-Attachment. It detects protein-complex core comprising of continually expressed and highly connected proteins in dynamic PPI network, and then the protein complex is formed by including the attachments with high adhesion into the core. The integration of core-attachment feature into the dynamic PPI network is responsible for the superiority of our algorithm. DCA has been applied on two different yeast dynamic PPI networks and the experimental results show that it performs significantly better than the state-of-the-art techniques in terms of prediction accuracy, hF-measure and statistical significance in biology. In addition, the identified complexes with strong biological significance provide potential candidate complexes for biologists to validate.

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

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)

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

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.

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)

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)

Identifying Core Competencies to Advance Female Professors' Careers: An Exploratory Study in United States Academia

Science.gov (United States)

Seo, Ga-eun; Hedayati Mehdiabadi, Amir; Huang, Wenhao

2017-01-01

This exploratory study aims to identify the core competencies necessary to successfully advance the careers of female associate professors in higher education. To ascertain these core career competencies, a critical incident interview technique was employed. One-to-one semi-structured interviews with six female full professors at a major research…

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

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.

A High-Throughput Cell-Based Screen Identified a 2-[(E)-2-Phenylvinyl]-8-Quinolinol Core Structure That Activates p53.

Science.gov (United States)

Bechill, John; Zhong, Rong; Zhang, Chen; Solomaha, Elena; Spiotto, Michael T

2016-01-01

p53 function is frequently inhibited in cancer either through mutations or by increased degradation via MDM2 and/or E6AP E3-ubiquitin ligases. Most agents that restore p53 expression act by binding MDM2 or E6AP to prevent p53 degradation. However, fewer compounds directly bind to and activate p53. Here, we identified compounds that shared a core structure that bound p53, caused nuclear localization of p53 and caused cell death. To identify these compounds, we developed a novel cell-based screen to redirect p53 degradation to the Skip-Cullin-F-box (SCF) ubiquitin ligase complex in cells expressing high levels of p53. In a multiplexed assay, we coupled p53 targeted degradation with Rb1 targeted degradation in order to identify compounds that prevented p53 degradation while not inhibiting degradation through the SCF complex or other proteolytic machinery. High-throughput screening identified several leads that shared a common 2-[(E)-2-phenylvinyl]-8-quinolinol core structure that stabilized p53. Surface plasmon resonance analysis indicated that these compounds bound p53 with a KD of 200 ± 52 nM. Furthermore, these compounds increased p53 nuclear localization and transcription of the p53 target genes PUMA, BAX, p21 and FAS in cancer cells. Although p53-null cells had a 2.5±0.5-fold greater viability compared to p53 wild type cells after treatment with core compounds, loss of p53 did not completely rescue cell viability suggesting that compounds may target both p53-dependent and p53-independent pathways to inhibit cell proliferation. Thus, we present a novel, cell-based high-throughput screen to identify a 2-[(E)-2-phenylvinyl]-8-quinolinol core structure that bound to p53 and increased p53 activity in cancer cells. These compounds may serve as anti-neoplastic agents in part by targeting p53 as well as other potential pathways.

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.

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.

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

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.

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

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

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

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).

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

Promise and Pitfalls of Using Grain Size Analysis to Identify Glacial Sediments in Alpine Lake Cores.

Science.gov (United States)

Clark, D. H.

2011-12-01

Lakes fed by glacier outwash should have a clastic particle-size record distinct from non-glacial lakes in the same area, but do they? The unique turquoise color of alpine glacial lakes reflects the flux of suspended clastic glacial rock flour to those lakes; conversely, lakes not fed by outwash are generally clear with sediments dominated by organics or slope-wash from nearby hillslopes. This contrast in sediment types and sources should produce a distinct and measureable different in grain sizes between the two settings. Results from a variety of lakes suggest the actual situation is often more subtle and complex. I compare grain size results to other proxies to assess the value of grain size analysis for paleoglacier studies. Over the past 10 years, my colleagues and I have collected and analyzed sediment cores from a wide variety of lakes below small alpine glaciers in an attempt to constrain the timing and magnitude of alpine glaciation in those basins. The basic concept is that these lakes act as continuous catchments for any rock flour produced upstream by glacier abrasion; as a glacier grows, the flux of rock flour to the lake will also increase. If the glacier disappears entirely, rock flour deposition will also cease in short order. We have focused our research in basins with simple sedimentologic settings: mostly small, high-altitude, stripped granitic or metamorphic cirques in which the cirque glaciers are the primary source of clastic sediments. In most cases, the lakes are fed by meltwater from a modern glacier, but were ice free during the earlier Holocene. In such cases, the lake cores should record formation of and changes in activity of the glacier upstream. We used a Malvern Mastersizer 2000 laser particle size analyzer for our grain size analyses, as well as recording magnetic susceptibility, color, and organics for the same cores. The results indicate that although lakes often experience increases in silt and clay-size (<0.63 mm) clastic

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.

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)

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)

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

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)

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.

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)

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

Using Latent Semantic Analysis to Identify Research Trends in OpenStreetMap

Directory of Open Access Journals (Sweden)

Sukhjit Singh Sehra

2017-07-01

Full Text Available OpenStreetMap (OSM, based on collaborative mapping, has become a subject of great interest to the academic community, resulting in a considerable body of literature produced by many researchers. In this paper, we use Latent Semantic Analysis (LSA to help identify the emerging research trends in OSM. An extensive corpus of 485 academic abstracts of papers published during the period 2007–2016 was used. Five core research areas and fifty research trends were identified in this study. In addition, potential future research directions have been provided to aid geospatial information scientists, technologists and researchers in undertaking future OSM research.

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

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.

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)

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

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.

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.

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.

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)

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)

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

Diagnose Test-Taker's Profile in Terms of Core Profile Patterns: Principal Component (PC) vs. Profile Analysis via MDS (PAMS) Approaches.

Science.gov (United States)

Kim, Se-Kang; Davison, Mark L.

A study was conducted to examine how principal components analysis (PCA) and Profile Analysis via Multidimensional Scaling (PAMS) can be used to diagnose individuals observed score profiles in terms of core profile patterns identified by each method. The standardization sample from the Wechsler Intelligence Scale for Children, Third Edition…

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.

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

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.)

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

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)

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)

Use of the International Classification of Functioning, Disability and Health to identify preliminary comprehensive and brief core sets for Guillain Barre syndrome.

Science.gov (United States)

Khan, Fary; Pallant, Julie F

2011-01-01

To identify the preliminary comprehensive and brief core sets for Guillain Barre syndrome (GBS), in a Delphi process using the International Classification of Functioning, Disability and Health (ICF). Focus groups and a consensus process were used to identify ICF core sets for GBS. This included: preliminary ICF studies; empirical patient data collection for 77 GBS participants; review of the evidence base and treatment in GBS literature followed by a Delphi exercise with 23 physicians and allied health professionals in Melbourne, Australia. The expert consensus selected 99 second level ICF categories (in three rounds) which identify health domains relevant to GBS for multidisciplinary assessment. These domains were consistent with current practice and existing GBS literature. The comprehensive core set includes: 27 (23%) categories from the component 'body function', 7 (12%) categories from 'body structures', 43 (36%) from 'activities and participation' and 22 (29%) from the component 'environmental' factors. The brief set comprised 20 categories, 20% of categories in the comprehensive core set. The core set categories for GBS-related health need to be addressed in multidisciplinary care programs. Future clinical 'rating' of this set may facilitate scale development using the ICF in GBS. Further research is needed to confirm the generalisability of this set in clinical settings.

Textual Demands of Passages in Three English/Language Arts Common Core Assessments and One Core Literacy Program for Intermediate Elementary Grades

Science.gov (United States)

Morrison, Timothy G.; Wilcox, Brad; Murdoch, Erica; Bird, Lauren

2018-01-01

The Common Core has emphasized reading for comprehension, including making inferences. However, little is known about the textual demands found within assessment and instructional passages that are promoted as being in line with Common Core expectations. The purpose of this content analysis was to identify the readability levels, passage length,…

Linking Core Promoter Classes to Circadian Transcription.

Directory of Open Access Journals (Sweden)

Pål O Westermark

2016-08-01

Full Text Available Circadian rhythms in transcription are generated by rhythmic abundances and DNA binding activities of transcription factors. Propagation of rhythms to transcriptional initiation involves the core promoter, its chromatin state, and the basal transcription machinery. Here, I characterize core promoters and chromatin states of genes transcribed in a circadian manner in mouse liver and in Drosophila. It is shown that the core promoter is a critical determinant of circadian mRNA expression in both species. A distinct core promoter class, strong circadian promoters (SCPs, is identified in mouse liver but not Drosophila. SCPs are defined by specific core promoter features, and are shown to drive circadian transcriptional activities with both high averages and high amplitudes. Data analysis and mathematical modeling further provided evidence for rhythmic regulation of both polymerase II recruitment and pause release at SCPs. The analysis provides a comprehensive and systematic view of core promoters and their link to circadian mRNA expression in mouse and Drosophila, and thus reveals a crucial role for the core promoter in regulated, dynamic transcription.

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)

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)

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)

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.

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

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)

MRI-derived measurements of fibrous-cap and lipid-core thickness: the potential for identifying vulnerable carotid plaques in vivo

International Nuclear Information System (INIS)

Trivedi, Rikin A.; U-King-Im, Jean-Marie; Graves, Martin J.; Horsley, Jo; Goddard, Martin; Kirkpatrick, Peter J.; Gillard, Jonathan H.

2004-01-01

Vulnerable plaques have thin fibrous caps overlying large necrotic lipid cores. Recent studies have shown that high-resolution MR imaging can identify these components. We set out to determine whether in vivo high-resolution MRI could quantify this aspect of the vulnerable plaque. Forty consecutive patients scheduled for carotid endarterectomy underwent pre-operative in vivo multi-sequence MR imaging of the carotid artery. Individual plaque constituents were characterised on MR images. Fibrous-cap and lipid-core thickness was measured on MRI and histology images. Bland-Altman plots were generated to determine the level of agreement between the two methods. Multi-sequence MRI identified 133 corresponding MR and histology slices. Plaque calcification or haemorrhage was seen in 47 of these slices. MR and histology derived fibrous cap-lipid-core thickness ratios showed strong agreement with a mean difference between MR and histology ratios of 0.02 (±0.04). The intra-class correlation coefficient between two readers for measurements was 0.87 (95% confidence interval, 0.73 and 0.93). Multi-sequence, high-resolution MR imaging accurately quantified the relative thickness of fibrous-cap and lipid-core components of carotid atheromatous plaques. This may prove to be a useful tool to characterise vulnerable plaques in vivo. (orig.)

MRI-derived measurements of fibrous-cap and lipid-core thickness: the potential for identifying vulnerable carotid plaques in vivo

Energy Technology Data Exchange (ETDEWEB)

Trivedi, Rikin A. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Addenbrooke' s Hospital, Academic Department of Neurosurgery, Cambridge (United Kingdom); U-King-Im, Jean-Marie; Graves, Martin J. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Horsley, Jo; Goddard, Martin [Papworth Hospital, Department of Histopathology, Papworth Everard (United Kingdom); Kirkpatrick, Peter J. [Addenbrooke' s Hospital, Academic Department of Neurosurgery, Cambridge (United Kingdom); Gillard, Jonathan H. [Addenbrooke' s Hospital, University Department of Radiology, Cambridge (United Kingdom); Addenbrooke' s Hospital, Hills Road, Box 219, Cambridge (United Kingdom)

2004-09-01

Vulnerable plaques have thin fibrous caps overlying large necrotic lipid cores. Recent studies have shown that high-resolution MR imaging can identify these components. We set out to determine whether in vivo high-resolution MRI could quantify this aspect of the vulnerable plaque. Forty consecutive patients scheduled for carotid endarterectomy underwent pre-operative in vivo multi-sequence MR imaging of the carotid artery. Individual plaque constituents were characterised on MR images. Fibrous-cap and lipid-core thickness was measured on MRI and histology images. Bland-Altman plots were generated to determine the level of agreement between the two methods. Multi-sequence MRI identified 133 corresponding MR and histology slices. Plaque calcification or haemorrhage was seen in 47 of these slices. MR and histology derived fibrous cap-lipid-core thickness ratios showed strong agreement with a mean difference between MR and histology ratios of 0.02 ({+-}0.04). The intra-class correlation coefficient between two readers for measurements was 0.87 (95% confidence interval, 0.73 and 0.93). Multi-sequence, high-resolution MR imaging accurately quantified the relative thickness of fibrous-cap and lipid-core components of carotid atheromatous plaques. This may prove to be a useful tool to characterise vulnerable plaques in vivo. (orig.)

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 ...

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...

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)

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.

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.

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

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)

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)

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

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)

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

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

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

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

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.

Comparative genomics of 12 strains of Erwinia amylovora identifies a pan-genome with a large conserved core.

Directory of Open Access Journals (Sweden)

Rachel A Mann

Full Text Available The plant pathogen Erwinia amylovora can be divided into two host-specific groupings; strains infecting a broad range of hosts within the Rosaceae subfamily Spiraeoideae (e.g., Malus, Pyrus, Crataegus, Sorbus and strains infecting Rubus (raspberries and blackberries. Comparative genomic analysis of 12 strains representing distinct populations (e.g., geographic, temporal, host origin of E. amylovora was used to describe the pan-genome of this major pathogen. The pan-genome contains 5751 coding sequences and is highly conserved relative to other phytopathogenic bacteria comprising on average 89% conserved, core genes. The chromosomes of Spiraeoideae-infecting strains were highly homogeneous, while greater genetic diversity was observed between Spiraeoideae- and Rubus-infecting strains (and among individual Rubus-infecting strains, the majority of which was attributed to variable genomic islands. Based on genomic distance scores and phylogenetic analysis, the Rubus-infecting strain ATCC BAA-2158 was genetically more closely related to the Spiraeoideae-infecting strains of E. amylovora than it was to the other Rubus-infecting strains. Analysis of the accessory genomes of Spiraeoideae- and Rubus-infecting strains has identified putative host-specific determinants including variation in the effector protein HopX1(Ea and a putative secondary metabolite pathway only present in Rubus-infecting strains.

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

Tank 241-SY-101 push mode core sampling and analysis plan

International Nuclear Information System (INIS)

CONNER, J.M.

1998-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-SY-101 (SY-101). It is written in accordance with Data Quality Objective to Support Resolution of the Flammable Gas Safety Issue (Bauer 1998), Low Activity Waste Feed Data Quality Objectives (Wiemers and Miller 1997 and DOE 1998), Data Quality Objectives for TWRS Privatization Phase I: Confirm Tank T is an Appropriate Feed Source for Low-Activity Waste Feed Batch X (Certa 1998), and the Tank Safety Screening Data Quality Objective (Dukelow et al. 1995). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues apply to tank SY-101 for this sampling event. Brown et al. also identifies high-level waste, regulatory, pretreatment and disposal issues as applicable issues for this tank. However, these issues will not be addressed via this sampling event

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

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.

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

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)

The Expanded FindCore Method for Identification of a Core Atom Set for Assessment of Protein Structure Prediction

Science.gov (United States)

Snyder, David A.; Grullon, Jennifer; Huang, Yuanpeng J.; Tejero, Roberto; Montelione, Gaetano T.

2014-01-01

Maximizing the scientific impact of NMR-based structure determination requires robust and statistically sound methods for assessing the precision of NMR-derived structures. In particular, a method to define a core atom set for calculating superimpositions and validating structure predictions is critical to the use of NMR-derived structures as targets in the CASP competition. FindCore (D.A. Snyder and G.T. Montelione PROTEINS 2005;59:673–686) is a superimposition independent method for identifying a core atom set, and partitioning that set into domains. However, as FindCore optimizes superimposition by sensitively excluding not-well-defined atoms, the FindCore core may not comprise all atoms suitable for use in certain applications of NMR structures, including the CASP assessment process. Adapting the FindCore approach to assess predicted models against experimental NMR structures in CASP10 required modification of the FindCore method. This paper describes conventions and a standard protocol to calculate an “Expanded FindCore” atom set suitable for validation and application in biological and biophysical contexts. A key application of the Expanded FindCore method is to identify a core set of atoms in the experimental NMR structure for which it makes sense to validate predicted protein structure models. We demonstrate the application of this Expanded FindCore method in characterizing well-defined regions of 18 NMR-derived CASP10 target structures. The Expanded FindCore protocol defines “expanded core atom sets” that match an expert’s intuition of which parts of the structure are sufficiently well-defined to use in assessing CASP model predictions. We also illustrate the impact of this analysis on the CASP GDT assessment scores. PMID:24327305

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.)

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.

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.

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)

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.)

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

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.

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.

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)

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)

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

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)

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.

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.

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.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Energy Technology Data Exchange (ETDEWEB)

Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

2014-10-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

International Nuclear Information System (INIS)

Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

2014-01-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

Characterization Of Core Sample Collected From The Saltstone Disposal Facility

International Nuclear Information System (INIS)

Cozzi, A.; Duncan, A.

2010-01-01

During the month of September 2008, grout core samples were collected from the Saltstone Disposal Facility, Vault 4, cell E. This grout was placed during processing campaigns in December 2007 from Deliquification, Dissolution and Adjustment Batch 2 salt solution. The 4QCY07 Waste Acceptance Criteria sample collected on 11/16/07 represents the salt solution in the core samples. Core samples were retrieved to initiate the historical database of properties of emplaced Saltstone and to demonstrate the correlation between field collected and laboratory prepared samples. Three samples were collected from three different locations. Samples were collected using a two-inch diameter concrete coring bit. In April 2009, the core samples were removed from the evacuated sample container, inspected, transferred to PVC containers, and backfilled with nitrogen. Samples furthest from the wall were the most intact cylindrically shaped cored samples. The shade of the core samples darkened as the depth of coring increased. Based on the visual inspection, sample 3-3 was selected for all subsequent analysis. The density and porosity of the Vault 4 core sample, 1.90 g/cm 3 and 59.90% respectively, were comparable to values achieved for laboratory prepared samples. X-ray diffraction analysis identified phases consistent with the expectations for hydrated Saltstone. Microscopic analysis revealed morphology features characteristic of cementitious materials with fly ash and calcium silicate hydrate gel. When taken together, the results of the density, porosity, x-ray diffraction analysis and microscopic analysis support the conclusion that the Vault 4, Cell E core sample is representative of the expected waste form.

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.

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.

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

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

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

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)

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

Core BPEL

DEFF Research Database (Denmark)

Hallwyl, Tim; Højsgaard, Espen

The Web Services Business Process Execution Language (WS-BPEL) is a language for expressing business process behaviour based on web services. The language is intentionally not minimal but provides a rich set of constructs, allows omission of constructs by relying on defaults, and supports language......, does not allow omissions, and does not contain ignorable elements. We do so by identifying syntactic sugar, including default values, and ignorable elements in WS-BPEL. The analysis results in a translation from the full language to the core subset. Thus, we reduce the effort needed for working...

Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

International Nuclear Information System (INIS)

Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

2015-01-01

Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

Energy Technology Data Exchange (ETDEWEB)

Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2015-07-01

Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

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)

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

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)

Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

International Nuclear Information System (INIS)

Struwe, D.

1977-12-01

A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

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

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

A Genome-Wide Association Study and Complex Network Identify Four Core Hub Genes in Bipolar Disorder

Directory of Open Access Journals (Sweden)

Zengyan Xie

2017-12-01

Full Text Available Bipolar disorder is a common and severe mental illness with unsolved pathophysiology. A genome-wide association study (GWAS has been used to find a number of risk genes, but it is difficult for a GWAS to find genes indirectly associated with a disease. To find core hub genes, we introduce a network analysis after the GWAS was conducted. Six thousand four hundred fifty eight single nucleotide polymorphisms (SNPs with p < 0.01 were sifted out from Wellcome Trust Case Control Consortium (WTCCC dataset and mapped to 2045 genes, which are then compared with the protein–protein network. One hundred twelve genes with a degree >17 were chosen as hub genes from which five significant modules and four core hub genes (FBXL13, WDFY2, bFGF, and MTHFD1L were found. These core hub genes have not been reported to be directly associated with BD but may function by interacting with genes directly related to BD. Our method engenders new thoughts on finding genes indirectly associated with, but important for, complex diseases.

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)

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)

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

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

Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

International Nuclear Information System (INIS)

Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

1985-01-01

A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Conklin, J.C.

1986-08-01

The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

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)

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)

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

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.)

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

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)

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

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.

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

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.)

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

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.)

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.)

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

Genome-wide methylation analysis identifies a core set of hypermethylated genes in CIMP-H colorectal cancer.

Science.gov (United States)

McInnes, Tyler; Zou, Donghui; Rao, Dasari S; Munro, Francesca M; Phillips, Vicky L; McCall, John L; Black, Michael A; Reeve, Anthony E; Guilford, Parry J

2017-03-28

Aberrant DNA methylation profiles are a characteristic of all known cancer types, epitomized by the CpG island methylator phenotype (CIMP) in colorectal cancer (CRC). Hypermethylation has been observed at CpG islands throughout the genome, but it is unclear which factors determine whether an individual island becomes methylated in cancer. DNA methylation in CRC was analysed using the Illumina HumanMethylation450K array. Differentially methylated loci were identified using Significance Analysis of Microarrays (SAM) and the Wilcoxon Signed Rank (WSR) test. Unsupervised hierarchical clustering was used to identify methylation subtypes in CRC. In this study we characterized the DNA methylation profiles of 94 CRC tissues and their matched normal counterparts. Consistent with previous studies, unsupervized hierarchical clustering of genome-wide methylation data identified three subtypes within the tumour samples, designated CIMP-H, CIMP-L and CIMP-N, that showed high, low and very low methylation levels, respectively. Differential methylation between normal and tumour samples was analysed at the individual CpG level, and at the gene level. The distribution of hypermethylation in CIMP-N tumours showed high inter-tumour variability and appeared to be highly stochastic in nature, whereas CIMP-H tumours exhibited consistent hypermethylation at a subset of genes, in addition to a highly variable background of hypermethylated genes. EYA4, TFPI2 and TLX1 were hypermethylated in more than 90% of all tumours examined. One-hundred thirty-two genes were hypermethylated in 100% of CIMP-H tumours studied and these were highly enriched for functions relating to skeletal system development (Bonferroni adjusted p value =2.88E-15), segment specification (adjusted p value =9.62E-11), embryonic development (adjusted p value =1.52E-04), mesoderm development (adjusted p value =1.14E-20), and ectoderm development (adjusted p value =7.94E-16). Our genome-wide characterization of DNA

Intestinal microbiota in healthy adults: temporal analysis reveals individual and common core and relation to intestinal symptoms.

Directory of Open Access Journals (Sweden)

Jonna Jalanka-Tuovinen

Full Text Available While our knowledge of the intestinal microbiota during disease is accumulating, basic information of the microbiota in healthy subjects is still scarce. The aim of this study was to characterize the intestinal microbiota of healthy adults and specifically address its temporal stability, core microbiota and relation with intestinal symptoms. We carried out a longitudinal study by following a set of 15 healthy Finnish subjects for seven weeks and regularly assessed their intestinal bacteria and archaea with the Human Intestinal Tract (HIT Chip, a phylogenetic microarray, in conjunction with qPCR analyses. The health perception and occurrence of intestinal symptoms was recorded by questionnaire at each sampling point.A high overall temporal stability of the microbiota was observed. Five subjects showed transient microbiota destabilization, which correlated not only with the intake of antibiotics but also with overseas travelling and temporary illness, expanding the hitherto known factors affecting the intestinal microbiota. We identified significant correlations between the microbiota and common intestinal symptoms, including abdominal pain and bloating. The most striking finding was the inverse correlation between Bifidobacteria and abdominal pain: subjects who experienced pain had over five-fold less Bifidobacteria compared to those without pain. Finally, a novel computational approach was used to define the common core microbiota, highlighting the role of the analysis depth in finding the phylogenetic core and estimating its size. The in-depth analysis suggested that we share a substantial number of our intestinal phylotypes but as they represent highly variable proportions of the total community, many of them often remain undetected.A global and high-resolution microbiota analysis was carried out to determine the temporal stability, the associations with intestinal symptoms, and the individual and common core microbiota in healthy adults. The

Tank 241-SX-105 rotary mode core sampling and analysis plan

International Nuclear Information System (INIS)

Simpson, B.C.

1998-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for rotary mode core samples from tank 241-SX-105 (SX-105). It is written in accordance with Tank Safety Screening Data Quality Objective (Dukelow et al. 1995) and Memorandum of Understanding for the Organic Complexant Safety Issue Data Requirements (Schreiber 1997a). Vapor screening issues apply as well, but are outside the scope of this SAP. A physical profile prediction based on waste fill history and previous sampling information is provided in Appendix A. Prior to core sampling, the dome space (below the riser) shall be measured for the presence of flammable gases. The measurement shall be taken from within the dome space and the data reported as a percentage of the lower flammability limit (LFL). The results shall be transmitted to the tank coordinator within ten working days of the sampling event (Schreiber 1997b). If the results are above 25 percent of the LFL when analyzing by gas chromatography/mass spectrometry or gas-specific monitoring gauges or above 10% of the LFL when analyzing with a combustible gas meter, the necessity for recurring sampling for flammable gas concentration and the frequency of such sampling will be determined by the Flammable Gas Safety Project. Any additional vapor sampling is not within the scope of this SAP

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

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.

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/.

Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

International Nuclear Information System (INIS)

Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.

2008-01-01

Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

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 ...

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

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

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)

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)

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)

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

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.

Managing water addition to a degraded core

International Nuclear Information System (INIS)

Kuan, P.; Hanson, D.J.; Odar, F.

1992-01-01

In this paper the authors present information that can be used in severe accident management by providing an improved understanding of the effects of water addition to a degraded core. This improved understanding is developed using a diagram showing a sequence of core damage states. Whenever possible, a temperature and a time after accident initiation are estimated for each damage state in the sequence diagram. This diagram can be used to anticipate the evolution of events during an accident. Possible responses of plant instruments are described to identify these damage states and the effects of water addition. The rate and amount of water addition needed (a) to remove energy from the core, (b) to stabilize the core or (c) to not adversely affect the damage progression, are estimated. Analysis of the capability to remove energy from large cohesive and particulate debris beds indicates that these beds may not be stabilized in the core region and they may partially relocate to the lower plenum of the reactor vessel

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

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)

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)

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

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.

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

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

Comparative genomic analysis of Helicobacter pylori from Malaysia identifies three distinct lineages suggestive of differential evolution.

Science.gov (United States)

Kumar, Narender; Mariappan, Vanitha; Baddam, Ramani; Lankapalli, Aditya K; Shaik, Sabiha; Goh, Khean-Lee; Loke, Mun Fai; Perkins, Tim; Benghezal, Mohammed; Hasnain, Seyed E; Vadivelu, Jamuna; Marshall, Barry J; Ahmed, Niyaz

2015-01-01

The discordant prevalence of Helicobacter pylori and its related diseases, for a long time, fostered certain enigmatic situations observed in the countries of the southern world. Variation in H. pylori infection rates and disease outcomes among different populations in multi-ethnic Malaysia provides a unique opportunity to understand dynamics of host-pathogen interaction and genome evolution. In this study, we extensively analyzed and compared genomes of 27 Malaysian H. pylori isolates and identified three major phylogeographic lineages: hspEastAsia, hpEurope and hpSouthIndia. The analysis of the virulence genes within the core genome, however, revealed a comparable pathogenic potential of the strains. In addition, we identified four genes limited to strains of East-Asian lineage. Our analyses identified a few strain-specific genes encoding restriction modification systems and outlined 311 core genes possibly under differential evolutionary constraints, among the strains representing different ethnic groups. The cagA and vacA genes also showed variations in accordance with the host genetic background of the strains. Moreover, restriction modification genes were found to be significantly enriched in East-Asian strains. An understanding of these variations in the genome content would provide significant insights into various adaptive and host modulation strategies harnessed by H. pylori to effectively persist in a host-specific manner. © The Author(s) 2014. Published by Oxford University Press on behalf of Nucleic Acids Research.

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

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...

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

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

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

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

THE JCMT GOULD BELT SURVEY: DENSE CORE CLUSTERS IN ORION A

International Nuclear Information System (INIS)

Lane, J.; Kirk, H.; Johnstone, D.; Mairs, S.; Francesco, J. Di; Sadavoy, S.; Hatchell, J.; Berry, D. S.; Jenness, T.; Hogerheijde, M. R.; Ward-Thompson, D.

2016-01-01

The Orion A molecular cloud is one of the most well-studied nearby star-forming regions, and includes regions of both highly clustered and more dispersed star formation across its full extent. Here, we analyze dense, star-forming cores identified in the 850 and 450 μ m SCUBA-2 maps from the JCMT Gould Belt Legacy Survey. We identify dense cores in a uniform manner across the Orion A cloud and analyze their clustering properties. Using two independent lines of analysis, we find evidence that clusters of dense cores tend to be mass segregated, suggesting that stellar clusters may have some amount of primordial mass segregation already imprinted in them at an early stage. We also demonstrate that the dense core clusters have a tendency to be elongated, perhaps indicating a formation mechanism linked to the filamentary structure within molecular clouds.

THE JCMT GOULD BELT SURVEY: DENSE CORE CLUSTERS IN ORION A

Energy Technology Data Exchange (ETDEWEB)

Lane, J.; Kirk, H.; Johnstone, D.; Mairs, S.; Francesco, J. Di [NRC Herzberg Astronomy and Astrophysics, 5071 West Saanich Road, Victoria, BC, V9E 2E7 (Canada); Sadavoy, S. [Max Planck Institute for Astronomy, Königstuhl 17, D-69117 Heidelberg (Germany); Hatchell, J. [Physics and Astronomy, University of Exeter, Stocker Road, Exeter EX4 4QL (United Kingdom); Berry, D. S. [East Asian Observatory, 660 N. A‘ohōkū Place, University Park, Hilo, Hawaii 96720 (United States); Jenness, T. [Joint Astronomy Centre, 660 N. A‘ohōkū Place, University Park, Hilo, Hawaii 96720 (United States); Hogerheijde, M. R. [Leiden Observatory, Leiden University, P.O. Box 9513, 2300 RA Leiden (Netherlands); Ward-Thompson, D. [Jeremiah Horrocks Institute, University of Central Lancashire, Preston, Lancashire, PR1 2HE (United Kingdom); Collaboration: JCMT Gould Belt Survey Team

2016-12-10

The Orion A molecular cloud is one of the most well-studied nearby star-forming regions, and includes regions of both highly clustered and more dispersed star formation across its full extent. Here, we analyze dense, star-forming cores identified in the 850 and 450 μ m SCUBA-2 maps from the JCMT Gould Belt Legacy Survey. We identify dense cores in a uniform manner across the Orion A cloud and analyze their clustering properties. Using two independent lines of analysis, we find evidence that clusters of dense cores tend to be mass segregated, suggesting that stellar clusters may have some amount of primordial mass segregation already imprinted in them at an early stage. We also demonstrate that the dense core clusters have a tendency to be elongated, perhaps indicating a formation mechanism linked to the filamentary structure within molecular clouds.

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)

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...

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

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.

Towards an ICF Core Set for chronic musculoskeletal conditions: commonalities across ICF Core Sets for osteoarthritis, rheumatoid arthritis, osteoporosis, low back pain and chronic widespread pain.

Science.gov (United States)

Schwarzkopf, S R; Ewert, T; Dreinhöfer, K E; Cieza, A; Stucki, G

2008-11-01

The objective of the study was to identify commonalities among the International Classification of Functioning, Disability and Health (ICF) Core Sets of osteoarthritis (OA), osteoporosis (OP), low back pain (LBP), rheumatoid arthritis (RA) and chronic widespread pain (CWP). The aim is to identify relevant categories for the development of a tentative ICF Core Set for musculoskeletal and pain conditions. The ICF categories common to the five musculoskeletal and pain conditions in the Brief and Comprehensive ICF Core Sets were identified in three steps. In a first step, the commonalities across the Brief and Comprehensive ICF Core Sets for these conditions were examined. In a second and third step, we analysed the increase in commonalities when iteratively excluding one or two of the five conditions. In the first step, 29 common categories out of the total number of 120 categories were identified across the Comprehensive ICF Core Sets of all musculoskeletal and pain conditions, primarily in the component activities and participation. In the second and third step, we found that the exclusion of CWP across the Comprehensive ICF Core Sets increased the commonalities of the remaining four musculoskeletal conditions in a maximum of ten additional categories. The Brief ICF Core Sets of all musculoskeletal and pain conditions contain four common categories out of a total number of 62 categories. The iterative exclusion of a singular condition did not significantly increase the commonalities in the remaining. Based on our analysis, it seems possible to develop a tentative Comprehensive ICF Core Set across a number of musculoskeletal conditions including LBP, OA, OP and RA. However, the profile of functioning in people with CWP differs considerably and should not be further considered for a common ICF Core Set.

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)

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)

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...

Comparative analysis of unprotected loss-of-flow accidents for the 1.0 m EFR-LVC core using different computer codes

International Nuclear Information System (INIS)

Royl, P.; Frizonnet, J.M.; Moran, J.

1993-02-01

A comparative analysis of the unprotected loss of flow (ULOF) accident has been performed for the LVC core (Lower Void Core) of the European Fast Reactor EFR with the FRAX5B and FRAX5C codes from the AEA-T, the PHYSURAC code from CEA and the SAS4A REF92 code system developed jointly between KfK, CEA and PNC. The accident is triggered by the run down of the coolant pumps with failure to trip the reactor by the primary and/or secondary shutdown system. Only a limited amount of mitigating reactivity from the third shutdown line was considered so that the accident can progress into boiling and core disruption. This code outlines the important modelling differences and compares the different simulations. The discussion of the rather wide spectrum of calculated accident progressions identifies the generic differences, relates them to the applied models, and summarizes the key points that are responsible for the different progressions. A comparison of the consequence spectrum from all simulations indicates zero work energies for the majority of the calculations. All simulations show up the need for a continued accident analysis into the early and late transition phase

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.

Genome-Wide Temporal Expression Profiling in Caenorhabditis elegans Identifies a Core Gene Set Related to Long-Term Memory.

Science.gov (United States)

Freytag, Virginie; Probst, Sabine; Hadziselimovic, Nils; Boglari, Csaba; Hauser, Yannick; Peter, Fabian; Gabor Fenyves, Bank; Milnik, Annette; Demougin, Philippe; Vukojevic, Vanja; de Quervain, Dominique J-F; Papassotiropoulos, Andreas; Stetak, Attila

2017-07-12

The identification of genes related to encoding, storage, and retrieval of memories is a major interest in neuroscience. In the current study, we analyzed the temporal gene expression changes in a neuronal mRNA pool during an olfactory long-term associative memory (LTAM) in Caenorhabditis elegans hermaphrodites. Here, we identified a core set of 712 (538 upregulated and 174 downregulated) genes that follows three distinct temporal peaks demonstrating multiple gene regulation waves in LTAM. Compared with the previously published positive LTAM gene set (Lakhina et al., 2015), 50% of the identified upregulated genes here overlap with the previous dataset, possibly representing stimulus-independent memory-related genes. On the other hand, the remaining genes were not previously identified in positive associative memory and may specifically regulate aversive LTAM. Our results suggest a multistep gene activation process during the formation and retrieval of long-term memory and define general memory-implicated genes as well as conditioning-type-dependent gene sets. SIGNIFICANCE STATEMENT The identification of genes regulating different steps of memory is of major interest in neuroscience. Identification of common memory genes across different learning paradigms and the temporal activation of the genes are poorly studied. Here, we investigated the temporal aspects of Caenorhabditis elegans gene expression changes using aversive olfactory associative long-term memory (LTAM) and identified three major gene activation waves. Like in previous studies, aversive LTAM is also CREB dependent, and CREB activity is necessary immediately after training. Finally, we define a list of memory paradigm-independent core gene sets as well as conditioning-dependent genes. Copyright © 2017 the authors 0270-6474/17/376661-12$15.00/0.

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.

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

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

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)

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

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)

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.

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

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.

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.

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)

Using Multi-Core Systems for Rover Autonomy

Science.gov (United States)

Clement, Brad; Estlin, Tara; Bornstein, Benjamin; Springer, Paul; Anderson, Robert C.

2010-01-01

Task Objectives are: (1) Develop and demonstrate key capabilities for rover long-range science operations using multi-core computing, (a) Adapt three rover technologies to execute on SOA multi-core processor (b) Illustrate performance improvements achieved (c) Demonstrate adapted capabilities with rover hardware, (2) Targeting three high-level autonomy technologies (a) Two for onboard data analysis (b) One for onboard command sequencing/planning, (3) Technologies identified as enabling for future missions, (4)Benefits will be measured along several metrics: (a) Execution time / Power requirements (b) Number of data products processed per unit time (c) Solution quality

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

Evaluation of core distortion in FBR

International Nuclear Information System (INIS)

Ikarimoto, I.; Tanaka, M.; Okubo, Y.

1984-01-01

The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

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)

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...

Pan-Genome Analysis of Human Gastric Pathogen H. pylori: Comparative Genomics and Pathogenomics Approaches to Identify Regions Associated with Pathogenicity and Prediction of Potential Core Therapeutic Targets

DEFF Research Database (Denmark)

Ali, Amjad; Naz, Anam; Soares, Siomar C.

2015-01-01

-genome approach; the predicted conserved gene families (1,193) constitute similar to 77% of the average H. pylori genome and 45% of the global gene repertoire of the species. Reverse vaccinology strategies have been adopted to identify and narrow down the potential core-immunogenic candidates. Total of 28 nonhost....... Pan-genome analyses of the global representative H. pylori isolates consisting of 39 complete genomes are presented in this paper. Phylogenetic analyses have revealed close relationships among geographically diverse strains of H. pylori. The conservation among these genomes was further analyzed by pan...

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)

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.

Mangrove sediment core analysis of foraminiferal assemblages - a study at two sites along the western coast of India

Directory of Open Access Journals (Sweden)

P. Vidya

2014-02-01

Full Text Available Mangroves are an unique habitat and are largely influenced by sea level changes and wave energy. Foraminifera (Protista preserved in mangrove sediments provide an excellent proxy for deducing past conditions. One meter deep mangrove core samples at two sites on the western coast of India were collected. The foraminiferal assemblages at various depths showed significant changes in the abundance and diversity down the cores. A total of 59 species belonging to 32 genera, 24 families and five suborders were identified from the cores of these two sites. The cores showed an abundance of genus Rotalidium particularly the species Rotalidium annectans. Other species identified include Ammonia, Elphidium, Nonion, Spiroloculina, Quinqueloculina, Globigerinoides, etc. The pH, organic matter and CaCO3 also showed variations down the cores. There was a lack of correlation between sediment characteristics and the abundance of foraminifera in the cores. The low diversity and differences in distribution of foraminifera compared to surface intertidal samples may be due to intense post depositional changes or anthropogenic disturbances. The mangrove ecology thus appears disturbed by various factors.

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

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.

Using the International Classification of Functioning, Disability, and Health to identify outcome domains for a core outcome set for aphasia: a comparison of stakeholder perspectives.

Science.gov (United States)

Wallace, Sarah J; Worrall, Linda; Rose, Tanya; Le Dorze, Guylaine

2017-11-12

This study synthesised the findings of three separate consensus processes exploring the perspectives of key stakeholder groups about important aphasia treatment outcomes. This process was conducted to generate recommendations for outcome domains to be included in a core outcome set for aphasia treatment trials. International Classification of Functioning, Disability, and Health codes were examined to identify where the groups of: (1) people with aphasia, (2) family members, (3) aphasia researchers, and (4) aphasia clinicians/managers, demonstrated congruence in their perspectives regarding important treatment outcomes. Codes were contextualized using qualitative data. Congruence across three or more stakeholder groups was evident for ICF chapters: Mental functions; Communication; and Services, systems, and policies. Quality of life was explicitly identified by clinicians/managers and researchers, while people with aphasia and their families identified outcomes known to be determinants of quality of life. Core aphasia outcomes include: language, emotional wellbeing, communication, patient-reported satisfaction with treatment and impact of treatment, and quality of life. International Classification of Functioning, Disability, and Health coding can be used to compare stakeholder perspectives and identify domains for core outcome sets. Pairing coding with qualitative data may ensure important nuances of meaning are retained. Implications for rehabilitation The outcomes measured in treatment research should be relevant to stakeholders and support health care decision making. Core outcome sets (agreed, minimum set of outcomes, and outcome measures) are increasingly being used to ensure the relevancy and consistency of the outcomes measured in treatment studies. Important aphasia treatment outcomes span all components of the International Classification of Functioning, Disability, and Health. Stakeholders demonstrated congruence in the identification of important

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.

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

Laboratory ultrasonic pulse velocity logging for determination of elastic properties from rock core

Science.gov (United States)

Blacklock, Natalie Erin

During the development of deep underground excavations spalling and rockbursting have been recognized as significant mechanisms of violent brittle failure. In order to predict whether violent brittle failure will occur, it is important to identify the location of stiffness transitions that are associated with geologic structure. One approach to identify the effect of geologic structures is to apply borehole geophysical tools ahead of the tunnel advance. Stiffness transitions can be identified using mechanical property analysis surveys that combine acoustic velocity and density data to calculate acoustic estimates of elastic moduli. However, logistical concerns arise since the approach must be conducted at the advancing tunnel face. As a result, borehole mechanical property analyses are rarely used. Within this context, laboratory ultrasonic pulse velocity testing has been proposed as a potential alternative to borehole mechanical property analysis since moving the analysis to the laboratory would remove logistical constraints and improve safety for the evaluators. In addition to the traditional method of conducting velocity testing along the core axis, two new methodologies for point-focused testing were developed across the core diameter, and indirectly along intact lengths of drill core. The indirect test procedure was implemented in a continuous ultrasonic velocity test program along 573m of drill core to identify key geologic structures that generated transitions in ultrasonic elastic moduli. The test program was successful at identifying the location of geologic contacts, igneous intrusions, faults and shear structures. Ultrasonic values of Young's modulus and bulk modulus were determined at locations of significant velocity transitions to examine the potential for energy storage and energy release. Comparison of results from different ultrasonic velocity test configurations determined that the indirect test configuration provided underestimates for values of

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

International Nuclear Information System (INIS)

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

1998-01-01

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

NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

Directory of Open Access Journals (Sweden)

Tukiran Surbakti

2016-03-01

Full Text Available Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. 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. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini

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

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

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

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

Evolution of - and Core-Dominated Lava Flows Using Scaling Analysis

Science.gov (United States)

Castruccio, A.; Rust, A.; Sparks, R. S.

2010-12-01

We investigated the front evolution of simple lava flows on a slope using scaling arguments. For the retarding force acting against gravity, we analyzed three different cases: a flow controlled by a Newtonian viscosity, a flow controlled by the yield strength of a diffusively growing crust and a flow controlled by its core yield strength. These models were tested using previously published data of front evolution and volume discharge of 10 lava flow eruptions from 6 different volcanoes. Our analysis suggests that for basaltic eruptions with high effusion rate and low crystal content, (Hawaiian eruptions), the best fit of the data is with a Newtonian viscosity. For basaltic eruptions with lower effusion rates (Etna eruptions) or long duration andesitic eruptions (Lonquimay eruption, Chile) the flow is controlled by the yield strength of a growing crust. Finally, for very high crystalline lavas (Colima, Santiaguito) the flow is controlled by its core yield strength. The order of magnitude of the viscosities from our analysis is in the same range as previous studies using field measurements on the same lavas. The yield strength values for the growing crust and core of the flow are similar and with an order of magnitude of 10^5 Pa. This number is similar to yield strength values found in lava domes by different authors. The consistency of yield strength ~10^5 Pa is because larger stresses cause fracturing of very crystalline magma, which drastically reduces its effective strength. Furthermore, we used a 2-D analysis of a Bingham fluid flow on a slope to conclude that, for lower yield strength values, the flow is controlled mainly by its plastic viscosity and the lava can be effectively modelled as Newtonian. Our analysis provides a simple tool to evaluate the main controlling forces in the evolution of a lava flow, as well as the magnitude of its rheological properties, for eruptions of different compositions and conditions and may be useful to predict the evolution of

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

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

International Nuclear Information System (INIS)

Fabbris, Olivier; Dardour, Saied; Blaise, Patrick; Ferrasse, Jean-Henry; Saez, Manuel

2016-01-01

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

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)

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.

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

Web-based Core Design System Development

International Nuclear Information System (INIS)

Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

2011-01-01

The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

A lock circuit for a multi-core processor

DEFF Research Database (Denmark)

2015-01-01

An integrated circuit comprising a multiple processor cores and a lock circuit that comprises a queue register with respective bits set or reset via respective, connections dedicated to respective processor cores, whereby the queue register identifies those among the multiple processor cores...... that are enqueued in the queue register. Furthermore, the integrated circuit comprises a current register and a selector circuit configured to select a processor core and identify that processor core by a value in the current register. A selected processor core is a prioritized processor core among the cores...... configured with an integrated circuit; and a silicon die configured with an integrated circuit....

Identifying organizational deficiencies through root-cause analysis

International Nuclear Information System (INIS)

Tuli, R.W.; Apostolakis, G.E.

1996-01-01

All nuclear power plants incorporate root-cause analysis as an instrument to help identify and isolate key factors judged to be of significance following an incident or accident. Identifying the principal deficiencies can become very difficult when the event involves not only human and machine interaction, but possibly the underlying safety and quality culture of the organization. The current state of root-cause analysis is to conclude the investigation after identifying human and/or hardware failures. In this work, root-cause analysis is taken one step further by examining plant work processes and organizational factors. This extension is considered significant to the success of the analysis, especially when management deficiency is believed to contribute to the incident. The results of root-cause analysis can be most effectively implemented if the organization, as a whole, wishes to improve the overall operation of the plant by preventing similar incidents from occurring again. The study adds to the existing root-cause analysis the ability to localize the causes of undesirable events and to focus on those problems hidden deeply within the work processes that are routinely followed in the operation and maintenance of the facility

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

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)

Development of new core competencies for Taiwanese Emergency Medical Technicians.

Science.gov (United States)

Chang, Yu-Tung; Tsai, Kuang-Chau; Williams, Brett

2018-01-01

Core competencies are considered the foundation for establishing Emergency Medical Technician (EMT) and paramedic curricula, and for ensuring performance standards in the delivery of prehospital care. This study surveyed EMT instructors and medical directors to identify the most desirable core competencies for all levels of EMTs in Taiwan. A principal components analysis with Varimax rotation was conducted. An online questionnaire was distributed to obtain perspectives of EMT instructors and medical directors on the most desirable core competencies for EMTs. The target population was EMT training-course instructors and medical directors of fire departments in Taiwan. The questionnaire comprised 61 competency items, and multiple-choice and open-ended questions were used to obtain respondents' perspectives of the Taiwanese EMT training and education system. The results identified three factors at EMT-1 and EMT-2 levels and five factors at the EMT-Paramedic level. The factors for EMT-1 and EMT-2 were similar, and those for EMT-Paramedics identified further comprehensive competence perspectives. The key factors that appear to influence the development of the Taiwanese Emergency Medical Services (EMS) education system are the attitude of authorities, the licensure system, and legislation. The findings present new core competencies for the Taiwanese EMT system and provide capacity to redesign curricula and reconsider roles for EMT-1 and EMT-2 technicians. At the EMT-Paramedic level, the findings demonstrate the importance of incorporating competency standards in the current skills-based curriculum. Moreover, the core-competencies gap that exists between Taiwanese EMT-1s, EMT-2s, and EMT-Paramedics and internationally recognized core competencies needs to be addressed. By identifying the key factors that potentially impact the development of the EMS education system, such as the attitude of authorities, the licensure system, and legislation, these findings will inform

Modeling and analysis of neutron noise from an ex-core detector at a pressurized water reactor

International Nuclear Information System (INIS)

Wood, R.T.; Perez, R.B.

1991-01-01

Two applications of a noise diagnostic methodology were performed using ex-core neutron detector data from a pressurized water reactor (PWR). A feedback dynamics model of the neutron power spectral density (PSD) was derived from a low-order whole-plant physical model made stochastic using the Langevin technique. From a functional fit to plant data, the response of the dynamic system to changes in important physical parameters was evaluated by a direct sensitivity analysis. In addition, changes in monitored spectra were related to changes in physical parameters and detection thresholds using common surveillance discriminants were determined. A resonance model was developed from perturbation theory to give the ex-core neutron detector response for small in-core mechanical motions in terms of a pole-strength factor, a resonance asymmetry (or skewness) factor, a vibration damping factor, and a frequency of vibration. The mechanical motion parameters for several resonances were determined by a functional fit of the model to plant data taken at various times during a fuel cycle and were tracked to determine trends that indicated vibrational changes of reactor internals. In addition, the resonance model gave the ability to separate the resonant components of the PSD after the parameters had been identified. As a result, the behavior of several vibration peaks were monitored over a fuel cycle. 9 refs., 6 figs., 1 tab

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.

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

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

Core antigen and circulating anti-core antibody in hepatitis B infection

Energy Technology Data Exchange (ETDEWEB)

Howard, C R; Zuckerman, A J [London School of Hygiene and Tropical Medicine (UK)

1977-02-01

Core antigen was obtained from the sera of persistent chronic carriers of hepatitis B virus by centrifugation and treatment with Nonidet P40 and 2-mercaptoethanol. The separated core antigen was radiolabelled and identified as a nucleoprotein structure of buoyant density 1.36 g/cm/sup 3/ and possessing an isoelectric point of 4.4. This material was employed in a radioimmnoassay procedure of high sensitivity for the detection of core antibody. In a series of sera from patients with acute type B hepatitis, core antibody was demonstrated 2 to 3 weeks after the onset of jaundice during the period of surface antigenaemia. The presence of core antibody may therefore provide an accurate serological marker for the detection of active or recent virus replication in future epidemiological studies of hepatitis B infection.

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)

Trace elements in a dated ice core from Antarctica

International Nuclear Information System (INIS)

Keshin, S.S.; Xudong Huang; Olmez, I.; Langway, C.C. Jr.

1992-01-01

Aerosol particles from both natural and anthropogenic sources are emitted into the atmosphere and transported by wind systems by various mechanisms. Once airborne, the particles, which contain various trace elements, accumulate on the earth's surface as either condensation nuclei or by dry fallout processes. In the polar regions, these particles are incorporated and deposited in snow layers in sequential time-unit increments. The trace analysis of elements contained in dated annual snow layers provides a measure of the elemental chemistry content of the atmosphere for the same time interval. A 164-m-deep, 10-cm-diam ice core was obtained at Byrd Station, Antarctica, in November 1989. Other physical and chemistry studies on this ice core have identified its detailed chronology in annual increments for the past 1360 yr. This study presents the results of the instrumental neutron activation analysis (INAA) measurements made on 26 individually dated samples of this core, selected between the 6.43- and 118.15-m depths

Evolution characteristics of the network core in the Facebook.

Directory of Open Access Journals (Sweden)

Jian-Guo Liu

Full Text Available Statistical properties of the static networks have been extensively studied. However, online social networks are evolving dynamically, understanding the evolving characteristics of the core is one of major concerns in online social networks. In this paper, we empirically investigate the evolving characteristics of the Facebook core. Firstly, we separate the Facebook-link(FL and Facebook-wall(FW datasets into 28 snapshots in terms of timestamps. By employing the k-core decomposition method to identify the core of each snapshot, we find that the core sizes of the FL and FW networks approximately contain about 672 and 373 nodes regardless of the exponential growth of the network sizes. Secondly, we analyze evolving topological properties of the core, including the k-core value, assortative coefficient, clustering coefficient and the average shortest path length. Empirical results show that nodes in the core are getting more interconnected in the evolving process. Thirdly, we investigate the life span of nodes belonging to the core. More than 50% nodes stay in the core for more than one year, and 19% nodes always stay in the core from the first snapshot. Finally, we analyze the connections between the core and the whole network, and find that nodes belonging to the core prefer to connect nodes with high k-core values, rather than the high degrees ones. This work could provide new insights into the online social network analysis.

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

Determination of heavy metals and other elements by Neutron Activation Analysis in sediment cores of laguna Mar Chiquita (Cordoba, Argentina)

International Nuclear Information System (INIS)

Larizzatti, Flavio Eduardo

2001-01-01

Laguna Mar Chiquita is one of the largest water bodies of South America. It is a big lake of saline waters, and its geographic localization is SOMS' S, 62 deg 30' W, about 150 km Northwest of Cordoba, Argentina. Due to its large variability of hydrological budget, surface and water levels produced periods of low stands (LLP) and high stands (HLP). This fluctuation of water level also produces substantial changes in the water salinity. The principal tributary of the Laguna Mar Chiquita is Dulce River and also receives water from two other rivers: Suquia and Xanaes. The Suquia River drains in a small satellite lake, the Laguna del Plata. The purpose of the present work was to investigate the sediment composition of the Laguna Mar Chiquita (2 sediment cores) and del Plata (one sediment core) by using Neutron Activation Analysis (NAA) technique. The three 60 cm long sediment cores, sliced each 2 cm, were analyzed and 26 elements were determined (As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, La, Lu, Na, Nd, Rb, Sb, Sc, Se, Sm, Ta, Tb, Th, U, Yb, Zn e Zr). Other complementary techniques were utilized: macro elements (Al, Ca, Fe, K, Mg, Mn, Na, P, Si, Ti) were determined by X-ray fluorescence, and the mineralogical composition of the sediments was determined by X-ray diffraction. The results obtained did not show any indication of anthropic contribution in the sediment composition, and concentration of the majority of the elements analyzed is uniform along the entire profile of the analyzed core. Statistical analysis of elemental concentrations (Cluster Analysis) reflects that in Laguna del Plata the fine fractions of the sediments is dominated by detrital minerals, while in the Laguna Mar Chiquita, the neo formed minerals are the principal components. In both lakes, it was possible to identify compositional variations in the sediment segments, which may correspond to temporal fluctuations in the sedimentation conditions. The statistical analysis associated to sedimentation

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

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.

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

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-09-01

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

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

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)

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)

Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

2009-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

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)

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.

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.

Cross-cohort analysis identifies a TEAD4 ↔ MYCN positive-feedback loop as the core regulatory element of high-risk neuroblastoma. | Office of Cancer Genomics

Science.gov (United States)

High-risk neuroblastomas show a paucity of recurrent somatic mutations at diagnosis. As a result, the molecular basis for this aggressive phenotype remains elusive. Recent progress in regulatory network analysis helped us elucidate disease-driving mechanisms downstream of genomic alterations, including recurrent chromosomal alterations. Our analysis identified three molecular subtypes of high-risk neuroblastomas, consistent with chromosomal alterations, and identified subtype-specific master regulator (MR) proteins that were conserved across independent cohorts.

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

Identifying deformation mechanisms in the NEEM ice core using EBSD measurements

Science.gov (United States)

Kuiper, Ernst-Jan; Weikusat, Ilka; Drury, Martyn R.; Pennock, Gill M.; de Winter, Matthijs D. A.

2015-04-01

Deformation of ice in continental sized ice sheets determines the flow behavior of ice towards the sea. Basal dislocation glide is assumed to be the dominant deformation mechanism in the creep deformation of natural ice, but non-basal glide is active as well. Knowledge of what types of deformation mechanisms are active in polar ice is critical in predicting the response of ice sheets in future warmer climates and its contribution to sea level rise, because the activity of deformation mechanisms depends critically on deformation conditions (such as temperature) as well as on the material properties (such as grain size). One of the methods to study the deformation mechanisms in natural materials is Electron Backscattered Diffraction (EBSD). We obtained ca. 50 EBSD maps of five different depths from a Greenlandic ice core (NEEM). The step size varied between 8 and 25 micron depending on the size of the deformation features. The size of the maps varied from 2000 to 10000 grid point. Indexing rates were up to 95%, partially by saving and reanalyzing the EBSP patterns. With this method we can characterize subgrain boundaries and determine the lattice rotation configurations of each individual subgrain. Combining these observations with arrangement/geometry of subgrain boundaries the dislocation types can be determined, which form these boundaries. Three main types of subgrain boundaries have been recognized in Antarctic (EDML) ice core¹². Here, we present the first results obtained from EBSD measurements performed on the NEEM ice core samples from the last glacial period, focusing on the relevance of dislocation activity of the possible slip systems. Preliminary results show that all three subgrain types, recognized in the EDML core, occur in the NEEM samples. In addition to the classical boundaries made up of basal dislocations, subgrain boundaries made of non-basal dislocations are also common. ¹Weikusat, I.; de Winter, D. A. M.; Pennock, G. M.; Hayles, M

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

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

Precise natural radiography of seafloor sedimentary cores by using imaging plate

International Nuclear Information System (INIS)

Sugihara, Makoto; Tsuchiya, Noriyoshi

2007-01-01

The radiation measurement for seafloor sedimentary cores was carried out by using the Imaging Plate (IP). The precise natural radiography was obtained with 48 hours (exposure time) in a shield box. Layered structures of seafloor sediments that have relatively low radioactivity like biogenic matters could be detected by autoradiography, and it was possible to identify optically unclear layers of sediments with differences of radiation characteristics. The detection limit of thickness of suspended particle layers like volcanic ashes might be sub millimeters. It was possible to identify the pinpoint position where radioactive mineral emitted α-rays at the surface of sample. Identification of pinpoint position could be possible suppose radioactive minerals that emits α-ray exposed at the surface of sample. Distribution of dose (PSL value of IP) was related to CT value (X-ray CT), which indicates dose of sedimentary core have relationships with density and atomic number of sediments. The natural radiography for seafloor sediments cores by using the Imaging Plate is considered to be a powerful new tool as a non-destructive core analysis. (author)

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

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-01-01

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

A Practical Core Loss Model for Filter Inductors of Power Electronic Converters

DEFF Research Database (Denmark)

Matsumori, Hiroaki; Shimizu, Toshihisa; Wang, Xiongfei

2018-01-01

This paper proposes a core loss model for filter inductors of power electronic converters. The model allows a computationally efficient analysis on the core loss of the inductor under the square voltage excitation and the premagnetization condition. First, the core loss of the filter inductor under...... buck chopper excitation is evaluated with the proposed model and compared with the conventional methods. The comparison shows that the proposed method results in a better core loss prediction under the premagnetized condition than that of conventional alternatives. Then, the core loss of the filter...... inductor with the pulsewidth modulated inverter excitation is evaluated, which shows that the proposed model not only accurately predicts the core loss but also identifies the hysteresis loss part. These results demonstrate that the approach can further be used for the development of magnetic materials...

How cores grow by pebble accretion. I. Direct core growth

Science.gov (United States)

Brouwers, M. G.; Vazan, A.; Ormel, C. W.

2018-03-01

Context. Planet formation by pebble accretion is an alternative to planetesimal-driven core accretion. In this scenario, planets grow by the accretion of cm- to m-sized pebbles instead of km-sized planetesimals. One of the main differences with planetesimal-driven core accretion is the increased thermal ablation experienced by pebbles. This can provide early enrichment to the planet's envelope, which influences its subsequent evolution and changes the process of core growth. Aims: We aim to predict core masses and envelope compositions of planets that form by pebble accretion and compare mass deposition of pebbles to planetesimals. Specifically, we calculate the core mass where pebbles completely evaporate and are absorbed before reaching the core, which signifies the end of direct core growth. Methods: We model the early growth of a protoplanet by calculating the structure of its envelope, taking into account the fate of impacting pebbles or planetesimals. The region where high-Z material can exist in vapor form is determined by the temperature-dependent vapor pressure. We include enrichment effects by locally modifying the mean molecular weight of the envelope. Results: In the pebble case, three phases of core growth can be identified. In the first phase (Mcore mixes outwards, slowing core growth. In the third phase (Mcore > 0.5M⊕), the high-Z inner region expands outwards, absorbing an increasing fraction of the ablated material as vapor. Rainout ends before the core mass reaches 0.6 M⊕, terminating direct core growth. In the case of icy H2O pebbles, this happens before 0.1 M⊕. Conclusions: Our results indicate that pebble accretion can directly form rocky cores up to only 0.6 M⊕, and is unable to form similarly sized icy cores. Subsequent core growth can proceed indirectly when the planet cools, provided it is able to retain its high-Z material.

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

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)

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

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

Installation of JMTR core management system

International Nuclear Information System (INIS)

Imaizumi, Tomomi; Ide, Hiroshi; Naka, Michihiro; Komukai, Bunsaku; Nagao, Yoshiharu

2013-01-01

In order to carry out the core management after the reoperation of JMTR quickly and accurately, the authors took up the Standard Reactor Analysis Code (SRAC) system and core management support programs that are operating in a general-purpose large computer and transferred them to PC (OS: Linux), and newly established a JMTR core management system. As for the core analysis, this measure enabled an increase in the processing speed from the check of core arrangement to the result display of nuclear restriction values to about 60 times, compared with the conventional method. It was confirmed that the differences of calculation results originated from the difference of internal display of computers, associated with the transfer of each analysis code from GS21-400 system to PC-Linux, were within practically allowable level. In the future, this system will be applied to the core analysis of JMTR, as well as to the preparation of operation plans. (A.O.)

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

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)

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

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)

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

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

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

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

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

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).

Core genome conservation of Staphylococcus haemolyticus limits sequence based population structure analysis.

Science.gov (United States)

Cavanagh, Jorunn Pauline; Klingenberg, Claus; Hanssen, Anne-Merethe; Fredheim, Elizabeth Aarag; Francois, Patrice; Schrenzel, Jacques; Flægstad, Trond; Sollid, Johanna Ericson

2012-06-01

The notoriously multi-resistant Staphylococcus haemolyticus is an emerging pathogen causing serious infections in immunocompromised patients. Defining the population structure is important to detect outbreaks and spread of antimicrobial resistant clones. Currently, the standard typing technique is pulsed-field gel electrophoresis (PFGE). In this study we describe novel molecular typing schemes for S. haemolyticus using multi locus sequence typing (MLST) and multi locus variable number of tandem repeats (VNTR) analysis. Seven housekeeping genes (MLST) and five VNTR loci (MLVF) were selected for the novel typing schemes. A panel of 45 human and veterinary S. haemolyticus isolates was investigated. The collection had diverse PFGE patterns (38 PFGE types) and was sampled over a 20 year-period from eight countries. MLST resolved 17 sequence types (Simpsons index of diversity [SID]=0.877) and MLVF resolved 14 repeat types (SID=0.831). We found a low sequence diversity. Phylogenetic analysis clustered the isolates in three (MLST) and one (MLVF) clonal complexes, respectively. Taken together, neither the MLST nor the MLVF scheme was suitable to resolve the population structure of this S. haemolyticus collection. Future MLVF and MLST schemes will benefit from addition of more variable core genome sequences identified by comparing different fully sequenced S. haemolyticus genomes. Copyright © 2012 Elsevier B.V. All rights reserved.

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

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)

Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/27/1999, During Sludge Core Removal

International Nuclear Information System (INIS)

VISWANATH, R.S.

1999-01-01

This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371/Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA(trademark) canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples were radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2

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.

Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

International Nuclear Information System (INIS)

2013-01-01

The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

Computed Tomography Scanning and Geophysical Measurements of Core from the Coldstream 1MH Well

Energy Technology Data Exchange (ETDEWEB)

Crandall, Dustin M.; Brown, Sarah; Moore, Johnathan E.; Mackey, Paige E.; Paronish, Thomas J.

2018-03-05

The computed tomography (CT) facilities and the Multi-Sensor Core Logger (MSCL) at the National Energy Technology Laboratory (NETL) Morgantown, West Virginia site were used to characterize core of the Marcellus Shale from a vertical well, the Coldstream 1MH Well in Clearfield County, PA. The core is comprised primarily of the Marcellus Shale from a depth of 7,002 to 7,176 ft.

The primary impetus of this work is a collaboration between West Virginia University (WVU) and NETL to characterize core from multiple wells to better understand the structure and variation of the Marcellus and Utica shale formations. As part of this effort, bulk scans of core were obtained from the Coldstream 1MH well, provided by the Energy Corporation of America (now Greylock Energy). This report, and the associated scans, provide detailed datasets not typically available from unconventional shales for analysis. The resultant datasets are presented in this report, and can be accessed from NETL's Energy Data eXchange (EDX) online system using the following link: https://edx.netl.doe.gov/dataset/coldstream-1mh-well.

All equipment and techniques used were non-destructive, enabling future examinations to be performed on these cores. None of the equipment used was suitable for direct visualization of the shale pore space, although fractures and discontinuities were detectable with the methods tested. Low resolution CT imagery with the NETL medical CT scanner was performed on the entire core. Qualitative analysis of the medical CT images, coupled with x-ray fluorescence (XRF), P-wave, and magnetic susceptibility measurements from the MSCL were useful in identifying zones of interest for more detailed analysis as well as fractured zones. En echelon fractures were observed at 7,100 ft and were CT scanned using NETL’s industrial CT scanner at higher resolution. The ability to quickly identify key areas for more detailed study with higher resolution will save time and

Stress analysis of shielded receiver lifting frame for core sampler truck number-sign 2

International Nuclear Information System (INIS)

Ziada, H.H.

1994-01-01

This analysis evaluates the structural design adequacy of the shielded receiver lifting frame (SRLF) for the rotary mode core sampler truck number 2 (RMCST number-sign 2). The analysis considers the loads expected during operation of the SRLF. Most of the existing welds were not in conformance with those specified on the drawings, H-2-91715 and -91716 (RHO 1988a and RHO 1988b). Stress analysts and engineers examined the configuration of the welds connecting the frame members of the SRLF and those connecting the SRLF to the drill rig. In comparison to those shown on the drawing, some of the actual welds appear stronger and others undersized. For example, the actual fillet welds completely encircle the junctures of members, although the drawings show some welds to be on two sides only. Attempts to find the original design calculations were unsuccessful. To resolve the nonconformance, the critical welds were identified by analysis and subsequently inspected to ensure they are as large or larger than the minimum is defined by weld leg size. A required weld size, as determined by stress analysis, of 0.1 inch or larger is considered to be critical. This size was selected because no existing welds were found to be less than 0.125 inch. Analysis results led to weld modifications to strengthen the SRLF. The weld modifications performed are described in WHC 1994

Materials interaction tests to identify base and coating materials for an enhanced in-vessel core catcher design

Energy Technology Data Exchange (ETDEWEB)

Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Swank, W.D. [Idaho National Engineering and Environmental Laboratory, Idaho Falls ID (United States); Cheung, F.B. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park PA (United States); Suh, K.Y. [Seoul National University, Department of Nuclear Engineering, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, Severe Accident Research Project, Taejon (Korea, Republic of)

2004-07-01

An enhanced in-vessel core catcher is being designed and evaluated, it must ensure In-Vessel Retention of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. Tests suggest that 2 coatings can provide adequate protection to a stainless steel core catcher: -) a 500 {mu}m thick zirconium dioxide coating over a 100-200 {mu}m Inconel 718 bond coating, and -) a 500 {mu}m thick magnesium zirconate coating.

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

A pressure core ultrasonic test system for on-board analysis of gas hydrate-bearing sediments under in situ pressures.

Science.gov (United States)

Yang, Lei; Zhou, Weihua; Xue, Kaihua; Wei, Rupeng; Ling, Zheng

2018-05-01

The enormous potential as an alternative energy resource has made natural gas hydrates a material of intense research interest. Their exploration and sample characterization require a quick and effective analysis of the hydrate-bearing cores recovered under in situ pressures. Here a novel Pressure Core Ultrasonic Test System (PCUTS) for on-board analysis of sediment cores containing gas hydrates at in situ pressures is presented. The PCUTS is designed to be compatible with an on-board pressure core transfer device and a long gravity-piston pressure-retained corer. It provides several advantages over laboratory core analysis including quick and non-destructive detection, in situ and successive acoustic property acquisition, and remission of sample storage and transportation. The design of the unique assembly units to ensure the in situ detection is demonstrated, involving the U-type protecting jackets, transducer precession device, and pressure stabilization system. The in situ P-wave velocity measurements make the detection of gas hydrate existence in the sediments possible on-board. Performance tests have verified the feasibility and sensitivity of the ultrasonic test unit, showing the dependence of P-wave velocity on gas hydrate saturation. The PCUTS has been successfully applied for analysis of natural samples containing gas hydrates recovered from the South China Sea. It is indicated that on-board P-wave measurements could provide a quick and effective understanding of the hydrate occurrence in natural samples, which can assist further resource exploration, assessment, and subsequent detailed core analysis.

A pressure core ultrasonic test system for on-board analysis of gas hydrate-bearing sediments under in situ pressures

Science.gov (United States)

Yang, Lei; Zhou, Weihua; Xue, Kaihua; Wei, Rupeng; Ling, Zheng

2018-05-01

The enormous potential as an alternative energy resource has made natural gas hydrates a material of intense research interest. Their exploration and sample characterization require a quick and effective analysis of the hydrate-bearing cores recovered under in situ pressures. Here a novel Pressure Core Ultrasonic Test System (PCUTS) for on-board analysis of sediment cores containing gas hydrates at in situ pressures is presented. The PCUTS is designed to be compatible with an on-board pressure core transfer device and a long gravity-piston pressure-retained corer. It provides several advantages over laboratory core analysis including quick and non-destructive detection, in situ and successive acoustic property acquisition, and remission of sample storage and transportation. The design of the unique assembly units to ensure the in situ detection is demonstrated, involving the U-type protecting jackets, transducer precession device, and pressure stabilization system. The in situ P-wave velocity measurements make the detection of gas hydrate existence in the sediments possible on-board. Performance tests have verified the feasibility and sensitivity of the ultrasonic test unit, showing the dependence of P-wave velocity on gas hydrate saturation. The PCUTS has been successfully applied for analysis of natural samples containing gas hydrates recovered from the South China Sea. It is indicated that on-board P-wave measurements could provide a quick and effective understanding of the hydrate occurrence in natural samples, which can assist further resource exploration, assessment, and subsequent detailed core analysis.

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

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

Estimating core inflation : the role of oil price shocks and imported inflation

OpenAIRE

Bjørnland, Hilde Christiane

1997-01-01

This paper calculates core inflation, by imposing long run restrictions on a structural vector autoregression (VAR) model containing the growth rate of output, inflation and oil prices. Core inflation is identified as that component in inflation that has no long run effect on output. No restrictions are placed on the response of output and inflation to the oil price shocks. The analysis is applied to Norway and the United Kingdom, both oil producing OECD countries. A model that ...

Application Service Providers (ASP Adoption in Core and Non-Core Functions

Directory of Open Access Journals (Sweden)

Aman Y.M. Chan

2009-10-01

Full Text Available With the further improvement in internet bandwidth, connection stability and data transmission security, a new wave of Application Service Providers (ASP is on his way. The recent booming on some models such as Software Application as Service (SaaS and On-Demand in 2008, has led to emergence of ASP model in core business functions. The traditional IS outsourcing covers the non-core business functions that are not critical to business performance and competitive advantages. Comparing with traditional IS outsourcing, ASP is a new phenomenon that can be considered as an emerging innovation as it covers both core and non-core business functions. Most of the executives do not comprehend the difference and similarity between traditional IS outsourcing and ASP mode. Hence, we propose to conduct a research so as to identify the determinants (cost benefit, gap in IS capability complementing the company's strategic goal, and trust to ASP's service and security level and moderating factors (management's attitude in ownership & control, and company aggressiveness of ASP adoption decision in both core and non-core business functions.

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.

Factors influencing disaster nursing core competencies of emergency nurses.

Science.gov (United States)

Park, Hye-Young; Kim, Ji-Soo

2017-10-01

Emergency nurses are expected to provide required nursing services by using their professional expertise to reduce the risk posed by disasters. Thus, emergency nurses' disaster nursing core competencies are essential for coping with disasters. The purpose of the study reported here was to identify factors influencing the disaster nursing core competencies of emergency nurses. A survey was conducted among 231 emergency nurses working in 12 hospitals in South Korea. Data were collected on disaster-related experience, attitude, knowledge, and disaster nursing core competencies by means of a questionnaire. In multiple regression analysis, disaster-related experience exerted the strongest influence on disaster nursing core competencies, followed by disaster-related knowledge. The explanatory power of these factors was 25.6%, which was statistically significant (F=12.189, pcompetencies of emergency nurses could be improved through education and training programs that enhance their disaster preparedness. The nursing profession needs to participate actively in the development of disaster nursing education and training programs. Copyright © 2017 Elsevier Inc. All rights reserved.

Tank 241-AP-105, cores 208, 209 and 210, analytical results for the final report

International Nuclear Information System (INIS)

Nuzum, J.L.

1997-01-01

This document is the final laboratory report for Tank 241-AP-105. Push mode core segments were removed from Risers 24 and 28 between July 2, 1997, and July 14, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-AP-105 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997) and Tank Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT), differential scanning calorimetry (DSC) analysis, or total organic carbon (TOC) analysis exceeded the notification limits as stated in TSAP and DQO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report. Appearance and Sample Handling Two cores, each consisting of four segments, were expected from Tank 241-AP-105. Three cores were sampled, and complete cores were not obtained. TSAP states core samples should be transported to the laboratory within three calendar days from the time each segment is removed from the tank. This requirement was not met for all cores. Attachment 1 illustrates subsamples generated in the laboratory for analysis and identifies their sources. This reference also relates tank farm identification numbers to their corresponding 222-S Laboratory sample numbers

Tank 241-AP-105, cores 208, 209 and 210, analytical results for the final report

Energy Technology Data Exchange (ETDEWEB)

Nuzum, J.L.

1997-10-24

This document is the final laboratory report for Tank 241-AP-105. Push mode core segments were removed from Risers 24 and 28 between July 2, 1997, and July 14, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-AP-105 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997) and Tank Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT), differential scanning calorimetry (DSC) analysis, or total organic carbon (TOC) analysis exceeded the notification limits as stated in TSAP and DQO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report. Appearance and Sample Handling Two cores, each consisting of four segments, were expected from Tank 241-AP-105. Three cores were sampled, and complete cores were not obtained. TSAP states core samples should be transported to the laboratory within three calendar days from the time each segment is removed from the tank. This requirement was not met for all cores. Attachment 1 illustrates subsamples generated in the laboratory for analysis and identifies their sources. This reference also relates tank farm identification numbers to their corresponding 222-S Laboratory sample numbers.

One Health Core Competency Domains

Directory of Open Access Journals (Sweden)

Rebekah Frankson

2016-09-01

Full Text Available The emergence of complex global challenges at the convergence of human, animal, and environmental health has catalyzed a movement supporting ‘One Health’ approaches. Despite recognition of the importance of One Health approaches to address these complex challenges, little effort has been directed at identifying the seminal knowledge, skills and attitudes necessary for individuals to successfully contribute to One Health efforts. Between 2008 and 2011, three groups independently embarked on separate initiatives to identify core competencies for professionals involved with One Health approaches. Core competencies were considered critically important for guiding curriculum development and continuing professional education as they describe the knowledge, skills and attitudes required to be effective. A workshop was convened in 2012 to synthesize the various strands of work on One Health competencies. Despite having different mandates, participants, and approaches, all of these initiatives identified similar core competency domains: management; communication and informatics; values and ethics; leadership; teams and collaboration; roles and responsibilities; and systems thinking. These core competency domains have been used to develop new continuing professional education programs for One Health professionals and help university curricula prepare new graduates to be able to contribute more effectively to One Health approaches.

One Health Core Competency Domains.

Science.gov (United States)

Frankson, Rebekah; Hueston, William; Christian, Kira; Olson, Debra; Lee, Mary; Valeri, Linda; Hyatt, Raymond; Annelli, Joseph; Rubin, Carol

2016-01-01

The emergence of complex global challenges at the convergence of human, animal, and environmental health has catalyzed a movement supporting "One Health" approaches. Despite recognition of the importance of One Health approaches to address these complex challenges, little effort has been directed at identifying the seminal knowledge, skills, and attitudes necessary for individuals to successfully contribute to One Health efforts. Between 2008 and 2011, three groups independently embarked on separate initiatives to identify core competencies for professionals involved with One Health approaches. Core competencies were considered critically important for guiding curriculum development and continuing professional education, as they describe the knowledge, skills, and attitudes required to be effective. A workshop was convened in 2012 to synthesize the various strands of work on One Health competencies. Despite having different mandates, participants, and approaches, all of these initiatives identified similar core competency domains: management; communication and informatics; values and ethics; leadership; teams and collaboration; roles and responsibilities; and systems thinking. These core competency domains have been used to develop new continuing professional education programs for One Health professionals and help university curricula prepare new graduates to be able to contribute more effectively to One Health approaches.

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

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)

Fuzzy logic based power-efficient real-time multi-core system

CERN Document Server

Ahmed, Jameel; Najam, Shaheryar; Najam, Zohaib

2017-01-01

This book focuses on identifying the performance challenges involved in computer architectures, optimal configuration settings and analysing their impact on the performance of multi-core architectures. Proposing a power and throughput-aware fuzzy-logic-based reconfiguration for Multi-Processor Systems on Chip (MPSoCs) in both simulation and real-time environments, it is divided into two major parts. The first part deals with the simulation-based power and throughput-aware fuzzy logic reconfiguration for multi-core architectures, presenting the results of a detailed analysis on the factors impacting the power consumption and performance of MPSoCs. In turn, the second part highlights the real-time implementation of fuzzy-logic-based power-efficient reconfigurable multi-core architectures for Intel and Leone3 processors. .

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)

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.

GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (I)-theory and application

International Nuclear Information System (INIS)

Kim, Do Heon; Choi, Hang Bok

2001-01-01

A generalized perturbation theory (GPT) program, GENOVA, has been developed for the purpose of various applications to Canadian deuterium uranium (CANDU) reactor physics analyses. GENOVA was written under the framework of CANDU physics design and analysis code, RFSP. A sensitivity method based on the GPT was implemented in GENOVA to estimate various sensitivity coefficients related to the movement of zone controller units (ZCUs) existing in the CANDU reactor. The numerical algorithm for the sensitivity method was verified by a simple 2 x 2 node problem. The capability of predicting ZCU levels upon a refueling perturbation was validated for a CANDU-6 reactor problem. The applicability of GENOVA to the CANDU-6 core physics analysis has been demonstrated with the optimum refueling simulation and the uncertainty analysis problems. For the optimum refueling simulation, an optimum channel selection strategy has been proposed, using the ZCU level predicted by GENOVA. The refueling simulation of a CANDU-6 natural uranium core has shown that the ZCU levels are successfully controlled within the operating range while the channel and bundle powers are satisfying the license limits. An uncertainty analysis has been performed for the fuel composition heterogeneity of a CANDU DUPIC core, using the sensitivity coefficients generated by GENOVA. The results have shown that the uncertainty of the core performance parameter can be reduced appreciably when the contents of the major fissile isotopes are tightly controlled. GENOVA code has been successfully explored to supplement the weak points of the current design and analysis code, such as the incapacity of performing an optimum refueling simulation and uncertainty analysis. The sample calculations have shown that GENOVA has strong potential to be used for CANDU core analysis combined with the current design and analysis code, RFSP, especially for the development of advanced CANDU fuels

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

Synthesis of Au@Ag core-shell nanocubes containing varying shaped cores and their localized surface plasmon resonances.

Science.gov (United States)

Gong, Jianxiao; Zhou, Fei; Li, Zhiyuan; Tang, Zhiyong

2012-06-19

We have synthesized Au@Ag core-shell nanocubes containing Au cores with varying shapes and sizes through modified seed-mediated methods. Bromide ions are found to be crucial in the epitaxial growth of Ag atoms onto Au cores and in the formation of the shell's cubic shape. The Au@Ag core-shell nanocubes exhibit very abundant and distinct localized surface plasmon resonance (LSPR) properties, which are core-shape and size-dependent. With the help of theoretical calculation, the physical origin and the resonance mode profile of each LSPR peak are identified and studied. The core-shell nanocrystals with varying shaped cores offer a new rich category for LSPR control through the plasmonic coupling effect between core and shell materials.

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

Power Production Analysis of the OE Buoy WEC for the CORES Project

DEFF Research Database (Denmark)

Lavelle, John; Kofoed, Jens Peter

This report describes the analysis performed on the OE Buoy for the CORES project by the wave energy group at Aalborg University, Denmark. OE Buoy is a type of Oscillating Water Column (OWC) wave energy converter as part of the CORES project. This type of device is one of the most developed...... to extract energy from the ocean (1). Typically, a Wells turbine is used for the Power Take Off (PTO) for OWCs. The Wells turbine has the advantage that it is self-rectifying – with the ability to operate with either direction of airflow, which changes during each cycle of the wave. This type of turbine...... which a total of 39 hours of power production data was collected. A data acquisition system was used to sample the sensors on board and the generator shaft power time-series data was used in the analysis here. A wave-rider buoy, located at the site of OE Buoy and operated by the Marine Institute Ireland...

Migration: a core public health ethics issue.

Science.gov (United States)

Wild, V; Dawson, A

2018-05-01

In this article, we outline the link between migration, public health and ethics. Discussing relevant arguments about migration from the perspective of public health and public health ethics. Critical review of theories and frameworks, case-based analysis and systematic identification and discussion of challenges. Migration is a core issue of public health ethics and must take a case-based approach: seeking to identify the specific ethical dimensions and vulnerabilities in each particular context. Public health as a practice, built upon the core value of justice, requires the protection and promotion of migrants' well-being (even if this produces tension with immigration services). Ethical analysis should take all phases of migration into account: before, during and after transit. We argue that migration policies, at least as they relate to migrants' well-being, should be founded upon a shared humanity, respect for human rights and on the idea that effective public health cannot and should not be confined within the borders and to the citizens of any host country. We make the case for migration to be seen as a core issue of public health ethics. Copyright © 2018 The Royal Society for Public Health. Published by Elsevier Ltd. All rights reserved.

Further HTGR core support structure reliability studies. Interim report No. 1

International Nuclear Information System (INIS)

Platus, D.L.

1976-01-01

Results of a continuing effort to investigate high temperature gas cooled reactor (HTGR) core support structure reliability are described. Graphite material and core support structure component physical, mechanical and strength properties required for the reliability analysis are identified. Also described are experimental and associated analytical techniques for determining the required properties, a procedure for determining number of tests required, properties that might be monitored by special surveillance of the core support structure to improve reliability predictions, and recommendations for further studies. Emphasis in the study is directed towards developing a basic understanding of graphite failure and strength degradation mechanisms; and validating analytical methods for predicting strength and strength degradation from basic material properties

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

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

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-12-21

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

International Nuclear Information System (INIS)

Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh; Banerjee, Sanjoy; Sohal, Manohar; Schultz, Richard; McEligot, Donald M.

2015-01-01

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

Citation analysis did not provide a reliable assessment of core outcome set uptake.

Science.gov (United States)

Barnes, Karen L; Kirkham, Jamie J; Clarke, Mike; Williamson, Paula R

2017-06-01

The aim of the study was to evaluate citation analysis as an approach to measuring core outcome set (COS) uptake, by assessing whether the number of citations for a COS report could be used as a surrogate measure of uptake of the COS by clinical trialists. Citation data were obtained for COS reports published before 2010 in five disease areas (systemic sclerosis, rheumatoid arthritis, eczema, sepsis and critical care, and female sexual dysfunction). Those publications identified as a report of a clinical trial were examined to identify whether or not all outcomes in the COS were measured in the trial. Clinical trials measuring the relevant COS made up a small proportion of the total number of citations for COS reports. Not all trials citing a COS report measured all the recommended outcomes. Some trials cited the COS reports for other reasons, including the definition of a condition or other trial design issues addressed by the COS report. Although citation data can be readily accessed, it should not be assumed that the citing of a COS report indicates that a trial has measured the recommended COS. Alternative methods for assessing COS uptake are needed. Copyright © 2017 The Authors. Published by Elsevier Inc. All rights reserved.

Comparative genomics defines the core genome of the growing N4-like phage genus and identifies N4-like Roseophage specific genes

Directory of Open Access Journals (Sweden)

Jacqueline Zoe-Munn Chan

2014-10-01

Full Text Available Two bacteriophages, RPP1 and RLP1, infecting members of the marine Roseobacter clade were isolated from seawater. Their linear genomes are 74.7 and 74.6 kb and encode 91 and 92 coding DNA sequences, respectively. Around 30% of these are homologous to genes found in Enterobacter phage N4. Comparative genomics of these two new Roseobacter phages and twenty-three other sequenced N4-like phages (three infecting members of the Roseobacter lineage and twenty infecting other Gammaproteobacteria revealed that N4-like phages share a core genome of 14 genes responsible for control of gene expression, replication and virion proteins. Phylogenetic analysis of these genes placed the five N4-like roseophages (RN4 into a distinct subclade. Analysis of the RN4 phage genomes revealed they share a further 19 genes of which nine are found exclusively in RN4 phages and four appear to have been acquired from their bacterial hosts. Proteomic analysis of the RPP1 and RLP1 virions identified a second structural module present in the RN4 phages similar to that found in the Pseudomonas N4-like phage LIT1. Searches of various metagenomic databases, included the GOS database, using CDS sequences from RPP1 suggests these phages are widely distributed in marine environments in particular in the open ocean environment.

Development of new core competencies for Taiwanese Emergency Medical Technicians

Directory of Open Access Journals (Sweden)

Chang YT

2018-03-01

Full Text Available Yu-Tung Chang,1,2 Kuang-Chau Tsai,2 Brett Williams1,3 1Department of Community Emergency Health and Paramedic Practice, Faculty of Medicine, Nursing and Health Sciences, Monash University, Frankston, VIC, Australia; 2Department of Emergency Medicine, Far Eastern Memorial Hospital, New Taipei City, Taiwan; 3Division of Paramedicine, University of Tasmania, Hobart, TAS, Australia Objectives: Core competencies are considered the foundation for establishing Emergency Medical Technician (EMT and paramedic curricula, and for ensuring performance standards in the delivery of prehospital care. This study surveyed EMT instructors and medical directors to identify the most desirable core competencies for all levels of EMTs in Taiwan. Methods: A principal components analysis with Varimax rotation was conducted. An online questionnaire was distributed to obtain perspectives of EMT instructors and medical directors on the most desirable core competencies for EMTs. The target population was EMT training-course instructors and medical directors of fire departments in Taiwan. The questionnaire comprised 61 competency items, and multiple-choice and open-ended questions were used to obtain respondents’ perspectives of the Taiwanese EMT training and education system. Results: The results identified three factors at EMT-1 and EMT-2 levels and five factors at the EMT-Paramedic level. The factors for EMT-1 and EMT-2 were similar, and those for EMT-Paramedics identified further comprehensive competence perspectives. The key factors that appear to influence the development of the Taiwanese Emergency Medical Services (EMS education system are the attitude of authorities, the licensure system, and legislation. Conclusion: The findings present new core competencies for the Taiwanese EMT system and provide capacity to redesign curricula and reconsider roles for EMT-1 and EMT-2 technicians. At the EMT-Paramedic level, the findings demonstrate the importance of

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

International Nuclear Information System (INIS)

Sugino, Kazuteru; Shono, Akira

2002-04-01

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

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....

The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

International Nuclear Information System (INIS)

Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

2013-01-01

The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

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

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.

Proteomic analysis of the increased stress tolerance of saccharomyces cerevisiae encapsulated in liquid core alginate-chitosan capsules.

Directory of Open Access Journals (Sweden)

Johan O Westman

Full Text Available Saccharomyces cerevisiae CBS8066 encapsulated in semi-permeable alginate or alginate-chitosan liquid core capsules have been shown to have an enhanced tolerance towards complex dilute-acid lignocellulose hydrolysates and the lignocellulose-derived inhibitor furfural, as well as towards high temperatures. The underlying molecular reasons for these effects have however not been elucidated. In this study we have investigated the response of the encapsulation on the proteome level in the yeast cells, in comparison with cells grown freely in suspension under otherwise similar conditions. The proteomic analysis was performed on whole cell protein extracts using nLC-MS/MS with TMT® labelling and 2-D DIGE. 842 and 52 proteins were identified using each method, respectively. The abundances of 213 proteins were significantly different between encapsulated and suspended cells, with good correlation between the fold change ratios obtained by the two methods for proteins identified in both. Encapsulation of the yeast caused an up-regulation of glucose-repressed proteins and of both general and starvation-specific stress responses, such as the trehalose biosynthesis pathway, and down-regulation of proteins linked to growth and protein synthesis. The encapsulation leads to a lack of nutrients for cells close to the core of the capsule due to mass transfer limitations. The triggering of the stress response may be beneficial for the cells in certain conditions, for example leading to the increased tolerance towards high temperatures and certain inhibitors.

Applicability of the diffusion and simplified P3 theories for BWR pin-by-pin core analysis

International Nuclear Information System (INIS)

Tada, Kenichi; Yamamoto, Akio; Kitamura, Yasunori; Yamane, Yoshihiro; Watanabe, Masato; Noda, Hiroshi

2007-01-01

The pin-by-pin fine mesh core calculation method is considered as a candidate of next-generation core calculation method for BWR. In this study, the diffusion and the simplified P 3 (SP 3 ) theories are applied to the pin-by-pin core analysis of BWR. Performances of the diffusion and the SP 3 theories for cell-homogeneous pin-by-pin fine mesh BWR core analysis are evaluated through comparison with cell-heterogeneous detailed transport calculation by the method of characteristics (MOC). In this study, two-dimensional, 2x2 multi-assemblies geometry is used to compare the prediction accuracies of the diffusion and the SP 3 theories. The 2x2 multi- assemblies geometry consists of two types of 9x9 UO 2 assembly that have two different enrichment splittings. To mitigate the cell-homogenization error, the SPH method is applied for the pin-by-pin fine mesh calculation. The SPH method is a technique that reproduces a result of heterogeneous calculation by that of homogeneous calculation. The calculation results indicated that diffusion theory shows larger discrepancy than that of SP 3 theory on pin-wise fission rates. Furthermore, the accuracy of the diffusion theory would not be sufficient for the pin-by-pin fine mesh calculation. In contrast to the diffusion theory, the SP 3 theory shows much better accuracy on pin wise fission rates. Therefore, if the SP 3 theory is applied, the accuracy of the pin-by-pin fine mesh BWR core analysis will be higher and will be sufficient for production calculation. (author)

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

International Nuclear Information System (INIS)

Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

1999-09-01

A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

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.

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)

Molten salt reactors: reactor cores

International Nuclear Information System (INIS)

1983-01-01

In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

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)

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.

Family and infant characteristics associated with timing of core and non-core food introduction in early childhood

OpenAIRE

Schrempft, Stephanie; van Jaarsveld, Cornelia H.M.; Fisher, Abigail; Wardle, Jane

2013-01-01

Objective To identify family and infant characteristics associated with timing of introduction of two food types: core foods (nutrient-dense) and non-core foods (nutrient-poor) in a population-based sample of mothers and infants. Method Participants were 1861 mothers and infants from the Gemini twin birth cohort (one child per family). Family and infant characteristics were assessed when the infants were around 8 months old. Timing of introducing core and non-core foods was assessed at 8 and ...

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)

Identification of the core ecosystem services and their spatial heterogeneity in Poyang Lake area

Institute of Scientific and Technical Information of China (English)

Nana SHI; Jinyan ZHAN; Feng WU; Jifu DU

2009-01-01

According to the ecosystem assessment framework developed by the Millennium Ecosystem Assessment (MA), this paper designs an evaluation system of ecosystem services in Poyang Lake area. On the basis of relevant variables disaggregated to 1 km grid using the gridded 1 km, this paper employs factor analysis to extract a number of factors which characterize the ecosystem services of Poyang Lake area. The extracted principal component are then represented onto 1 km ×1 km grids by spatial clustering analysis to recognize and identify the minimal but consistent mapping units for ecosystem services which can be used to delimit the boundaries of ecological service zones. The research identifies ten ecosystem service zones in Poyang Lake area according to the consistent principle of core ecosystem service unit.Four kinds of core ecosystem services including supporting function, provisioning function, regulating function and cultural function are identified and represented. The research results could provide both spatially and temporally valuable decision-making information for sustainable ecosystem management in the targeted area.

Management Education for Archivists, Information Managers, and Librarians: Is There a Global Core?

Science.gov (United States)

Evans, G. Edward

1984-01-01

Reports on an analysis of 35 American management textbooks that identifies 17 core concepts for management courses for information professionals. Each concept--accountability, accounting, authority, budgeting, change/innovation/creativity, communication, controlling, decision making, delegation and organization, ethics, fiscal management,…

Transcriptional differences between normal and glioma-derived glial progenitor cells identify a core set of dysregulated genes.

Science.gov (United States)

Auvergne, Romane M; Sim, Fraser J; Wang, Su; Chandler-Militello, Devin; Burch, Jaclyn; Al Fanek, Yazan; Davis, Danielle; Benraiss, Abdellatif; Walter, Kevin; Achanta, Pragathi; Johnson, Mahlon; Quinones-Hinojosa, Alfredo; Natesan, Sridaran; Ford, Heide L; Goldman, Steven A

2013-06-27

Glial progenitor cells (GPCs) are a potential source of malignant gliomas. We used A2B5-based sorting to extract tumorigenic GPCs from human gliomas spanning World Health Organization grades II-IV. Messenger RNA profiling identified a cohort of genes that distinguished A2B5+ glioma tumor progenitor cells (TPCs) from A2B5+ GPCs isolated from normal white matter. A core set of genes and pathways was substantially dysregulated in A2B5+ TPCs, which included the transcription factor SIX1 and its principal cofactors, EYA1 and DACH2. Small hairpin RNAi silencing of SIX1 inhibited the expansion of glioma TPCs in vitro and in vivo, suggesting a critical and unrecognized role of the SIX1-EYA1-DACH2 system in glioma genesis or progression. By comparing the expression patterns of glioma TPCs with those of normal GPCs, we have identified a discrete set of pathways by which glial tumorigenesis may be better understood and more specifically targeted. Copyright © 2013 The Authors. Published by Elsevier Inc. All rights reserved.

Robust Detection and Visualization of Jet-Stream Core Lines in Atmospheric Flow.

Science.gov (United States)

Kern, Michael; Hewson, Tim; Sadlo, Filip; Westermann, Rudiger; Rautenhaus, Marc

2018-01-01

Jet-streams, their core lines and their role in atmospheric dynamics have been subject to considerable meteorological research since the first half of the twentieth century. Yet, until today no consistent automated feature detection approach has been proposed to identify jet-stream core lines from 3D wind fields. Such 3D core lines can facilitate meteorological analyses previously not possible. Although jet-stream cores can be manually analyzed by meteorologists in 2D as height ridges in the wind speed field, to the best of our knowledge no automated ridge detection approach has been applied to jet-stream core detection. In this work, we -a team of visualization scientists and meteorologists-propose a method that exploits directional information in the wind field to extract core lines in a robust and numerically less involved manner than traditional 3D ridge detection. For the first time, we apply the extracted 3D core lines to meteorological analysis, considering real-world case studies and demonstrating our method's benefits for weather forecasting and meteorological research.

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.

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

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

2014-01-01

Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

Identifying and Eliminating Deficiencies in the General Surgery Resident Core Competency Curriculum.

Science.gov (United States)

Tapia, Nicole M; Milewicz, Allen; Whitney, Stephen E; Liang, Michael K; Braxton, Carla C

2014-06-01

Although the Accreditation Council for Graduate Medical Education has defined 6 core competencies required of resident education, no consensus exists on best practices for reaching resident proficiency. Surgery programs must develop resourceful methods to incorporate learning. While patient care and medical knowledge are approached with formal didactics and traditional Halstedian educational formats, other core competencies are presumed to be learned on the job or emphasized in conferences. To test the hypothesis that our residents lack a foundation in several of the nonclinical core competencies and to seek to develop a formal curriculum that can be integrated into our current didactic time, with minimal effect on resident work hours and rest hours. Anonymous Likert-type scale needs assessment survey requesting residents within a large single general surgery residency program to rate their understanding, working knowledge, or level of comfort on the following 10 topics: negotiation and conflict resolution; leadership styles; health care legislation; principles of quality delivery of care, patient safety, and performance improvement; business of medicine; clinical practice models; role of advocacy in health care policy and government; personal finance management; team building; and roles of innovation and technology in health care delivery. Proportions of resident responses scored as positive (agree or strongly agree) or negative (disagree or strongly disagree). In total, 48 surgery residents (70%) responded to the survey. Only 3 topics (leadership styles, team building, and roles of innovation and technology in health care delivery) had greater than 70% positive responses, while 2 topics (negotiation and conflict resolution and principles of quality delivery of care, patient safety, and performance improvement) had greater than 60% positive responses. The remaining topics had less than 40% positive responses, with the least positive responses on the topics

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…

Discordance between location of positive cores in biopsy and location of positive surgical margin following radical prostatectomy.

Science.gov (United States)

Kim, Ji Won; Park, Hyoung Keun; Kim, Hyeong Gon; Ham, Dong Yeub; Paick, Sung Hyun; Lho, Yong Soo; Choi, Woo Suk

2015-10-01

We compared location of positive cores in biopsy and location of positive surgical margin (PSM) following radical prostatectomy. This retrospective analysis included patients who were diagnosed as prostate cancer by standard 12-core transrectal ultrasonography guided prostate biopsy, and who have PSM after radical prostatectomy. After exclusion of number of biopsy cores location data, 46 patients with PSM were identified. Locations of PSM in pathologic specimen were reported as 6 difference sites (apex, base and lateral in both sides). Discordance of biopsy result and PSM was defined when no positive cores in biopsy was identified at the location of PSM. Most common location of PSM were right apex (n=21) and left apex (n=15). Multiple PSM was reported in 21 specimens (45.7%). In 32 specimens (69.6%) with PSM, one or more concordant positive biopsy cores were identified, but 14 specimens (28%) had no concordant biopsy cores at PSM location. When discordant rate was separated by locations of PSM, right apex PSM had highest rate of discordant (38%). The discordant group had significantly lower prostate volume and lower number of positive cores in biopsy than concordant group. This study showed that one fourth of PSM occurred at location where tumor was not detected at biopsy and that apex PSM had highest rate of discordant. Careful dissection to avoid PSM should be performed in every location, including where tumor was not identified in biopsy.

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)

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

RASMUSSEN, J.H.

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 required to satisfy the Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank TIS An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase 1: Confirm Tank TIS An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999) and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues, except the Equipment DQO apply to tank 241-AZ-102 for this sampling event. The Equipment DQO is applied for shear strength measurements of the solids segments only. Poppiti (1999) requires additional americium-241 analyses of the sludge segments. Brown et al. (1998) also identify safety screening, regulatory issues and provision of samples to the Privatization Contractor(s) as applicable issues for this tank. However, these issues will not be addressed via this sampling event. Reynolds et al. (1999) concluded that information from previous sampling events was sufficient to satisfy the safety screening requirements for tank 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

CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

International Nuclear Information System (INIS)

Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

2010-01-01

The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

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

Analysis Of The Effect Of Fuel Enrichment Error On Neutronic Properties Of The RSG-GAS Core

International Nuclear Information System (INIS)

Saragih, Tukiran; Pinem, Surian

2002-01-01

The analysis of the fuel enrichment error effect on neutronic properties has been carried out. The fuel enrichment could be improperly done because of wrong fabrication. Therefore it is necessary to analyze the fuel enrichment error effect to determine how many percents the fuel enrichment maximum can be accepted in the core. The analysis was done by simulation method The RSG-GAS core was simulated with 5 standard fuels and 1 control element having wrong enrichment when inserted into the core. Fuel enrichment error was then simulated from 20%, 25% and 30% and the simulation was done using WIMSD/4 and Batan-2DIFF codes. The cross section of core material of the RSG-GAS was generated by WIMSD/4 code in 1-D, X-Y geometry and 10 energy neutron group. Two dimensions, diffusion calculation based on finite element method was done by using Batan-2DIFF code. Five fuel elements and one control element changed the enrichment was finally arranged as a new core of the RSG-Gas reactor. The neutronic properties can be seen from eigenvalues (k eff ) as well as from the kinetic properties based on moderator void reactivity coefficient. The calculated results showed that the error are still acceptable by k eff 1,097 even until 25% fuel enrichment but not more than 25,5%

The study on the core personality trait words of Chinese medical university students based on social network analysis.

Science.gov (United States)

Wu, Ying; Xue, Yunzhen; Xue, Zhanling

2017-09-01

The medical university students in China whose school work is relatively heavy and educational system is long are a special professional group. Many students have psychological problems more or less. So, to understand their personality characteristics will provide a scientific basis for the intervention of psychological health.We selected top 30 personality trait words according to the order of frequency. Additionally, some methods such as social network analysis (SNA) and visualization technology of mapping knowledge domain were used in this study.Among these core personality trait words Family conscious had the 3 highest centralities and possessed the largest core status and influence. From the analysis of core-peripheral structure, we can see polarized core-perpheral structure was quite obvious. From the analysis of K-plex, there were in total 588 "K-2"K-plexs. From the analysis of Principal Components, we selected the 11 principal components.This study of personality not only can prevent disease, but also provide a scientific basis for students' psychological healthy education. In addition, we have adopted SNA to pay more attention to the relationship between personality trait words and the connection among personality dimensions. This study may provide the new ideas and methods for the research of personality structure.

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 ...

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

Final report on the sampling and analysis of sediment cores from the L-Area oil and chemical basin

Energy Technology Data Exchange (ETDEWEB)

1985-08-01

Nine vibracores were collected in the L-Area oil and chemical basin (904-83G) during late March and early April 1985. These cores were collected for analysis of the sludge on the basin floor and the underlying sediment. Several different field and laboratory analyses were performed on each three inch segment of all the cores. These included: (1) Sediment characterization; (2) Percent moisture; (3) Dry weight; (4) Spectral gamma analysis; (5) Gross alpha and beta analysis. Detailed chemical analysis were measured on selected intervals of 2 cores (LBC-5 and 6) for complete chemical characterization of the sediments. This sampling program was conducted to provide information so that a closure plan for the basin could be developed. This report describes the methods employed during the project and provide a hard copy of the analytical results from the sample analyses. Included in the appendices are copies of all field and laboratory notes taken during the project and copies of the gas chromatograms for the petroleum hydrocarbon analysis. All chemical results were also submitted on a 5-inch floppy disk.

The ICF Core Sets for hearing loss--researcher perspective. Part I: Systematic review of outcome measures identified in audiological research.

Science.gov (United States)

Granberg, Sarah; Dahlström, Jennie; Möller, Claes; Kähäri, Kim; Danermark, Berth

2014-02-01

To review the literature in order to identify outcome measures used in research on adults with hearing loss (HL) as part of the ICF Core Sets development project, and to describe study and population characteristics of the reviewed studies. A systematic review methodology was applied using multiple databases. A comprehensive search was conducted and two search pools were created, pool I and pool II. The study population included adults (≥ 18 years of age) with HL and oral language as the primary mode of communication. 122 studies were included. Outcome measures were distinguished by 'instrument type', and 10 types were identified. In total, 246 (pool I) and 122 (pool II) different measures were identified, and only approximately 20% were extracted twice or more. Most measures were related to speech recognition. Fifty-one different questionnaires were identified. Many studies used small sample sizes, and the sex of participants was not revealed in several studies. The low prevalence of identified measures reflects a lack of consensus regarding the optimal outcome measures to use in audiology. Reflections and discussions are made in relation to small sample sizes and the lack of sex differentiation/descriptions within the included articles.

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...

Core story creation: analysing narratives to construct stories for learning.

Science.gov (United States)

Petty, Julia; Jarvis, Joy; Thomas, Rebecca

2018-03-16

Educational research uses narrative enquiry to gain and interpret people's experiences. Narrative analysis is used to organise and make sense of acquired narrative. 'Core story creation' is a way of managing raw data obtained from narrative interviews to construct stories for learning. To explain how core story creation can be used to construct stories from raw narratives obtained by interviewing parents about their neonatal experiences and then use these stories to educate learners. Core story creation involves reconfiguration of raw narratives. Reconfiguration includes listening to and rereading transcribed narratives, identifying elements of 'emplotment' and reordering these to form a constructed story. Thematic analysis is then performed on the story to draw out learning themes informed by the participants. Core story creation using emplotment is a strategy of narrative reconfiguration that produces stories which can be used to develop resources relating to person-centred education about the patient experience. Stories constructed from raw narratives in the context of constructivism can provide a medium or an 'end product' for use in learning resource development. This can then contribute to educating students or health professionals about patients' experiences. ©2018 RCN Publishing Company Ltd. All rights reserved. Not to be copied, transmitted or recorded in any way, in whole or part, without prior permission of the publishers.

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

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...

Opportunities and Challenges of Linking Scientific Core Samples to the Geoscience Data Ecosystem

Science.gov (United States)

Noren, A. J.

2016-12-01

Core samples generated in scientific drilling and coring are critical for the advancement of the Earth Sciences. The scientific themes enabled by analysis of these samples are diverse, and include plate tectonics, ocean circulation, Earth-life system interactions (paleoclimate, paleobiology, paleoanthropology), Critical Zone processes, geothermal systems, deep biosphere, and many others, and substantial resources are invested in their collection and analysis. Linking core samples to researchers, datasets, publications, and funding agencies through registration of globally unique identifiers such as International Geo Sample Numbers (IGSNs) offers great potential for advancing several frontiers. These include maximizing sample discoverability, access, reuse, and return on investment; a means for credit to researchers; and documentation of project outputs to funding agencies. Thousands of kilometers of core samples and billions of derivative subsamples have been generated through thousands of investigators' projects, yet the vast majority of these samples are curated at only a small number of facilities. These numbers, combined with the substantial similarity in sample types, make core samples a compelling target for IGSN implementation. However, differences between core sample communities and other geoscience disciplines continue to create barriers to implementation. Core samples involve parent-child relationships spanning 8 or more generations, an exponential increase in sample numbers between levels in the hierarchy, concepts related to depth/position in the sample, requirements for associating data derived from core scanning and lithologic description with data derived from subsample analysis, and publications based on tens of thousands of co-registered scan data points and thousands of analyses of subsamples. These characteristics require specialized resources for accurate and consistent assignment of IGSNs, and a community of practice to establish norms

Bradford's law: identification of the core journals for neurosurgery and its subspecialties.

Science.gov (United States)

Venable, Garrett T; Shepherd, Brandon A; Loftis, Christopher M; McClatchy, S Gray; Roberts, Mallory L; Fillinger, Meghan E; Tansey, James B; Klimo, Paul

2016-02-01

Bradford's law describes the scatter of citations for a given subject or field. It can be used to identify the most highly cited journals for a field or subject. The objective of this study was to use currently accepted formulations of Bradford's law to identify core journals of neurosurgery and neurosurgical subspecialties. All original research publications from 2009 to 2013 were analyzed for the top 25 North American academic neurosurgeons from each subspecialty. The top 25 were chosen from a ranked career h-index list identified from previous studies. Egghe's formulation and the verbal formulation of Bradford's law were applied to create specific citation density zones and identify the core journals for each subspecialty. The databases were then combined to identify the core journals for all of academic neurosurgery. Using Bradford's verbal law with 4 zone models, the authors were able to identify the core journals of neurosurgery and its subspecialties. The journals found in the most highly cited first zone are presented here as the core journals. For neurosurgery as a whole, the core included the following journals: Journal of Neurosurgery, Neurosurgery, Spine, Stroke, Neurology, American Journal of Neuroradiology, International Journal of Radiation Oncology Biology Physics, and New England Journal of Medicine. The core journals for each subspecialty are presented in the manuscript. Bradford's law can be used to identify the core journals of neurosurgery and its subspecialties. The core journals vary for each neurosurgical subspecialty, but Journal of Neurosurgery and Neurosurgery are among the core journals for each neurosurgical subspecialty.

45-Day safety screening results for tank 241-U-102, push mode cores 143 and 144

International Nuclear Information System (INIS)

Steen, F.H.

1996-01-01

This document is the 45-day report deliverable for tank 241-U-102 push mode core segments collected between April 16, 1996 and May 6, 1996 and received by the 222-S Laboratory between April 17, 1996 and May 8, 1996. The segments were subsampled and analyzed in accordance, with the Tank 241-U-102 Push Mode Core Sampling and analysis Plan (TSAP) (Hu, 1996) and the Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). The analytical results are included in Table 1. Attachment I is a cross reference to relate the tank farm identification numbers to the 222-S Laboratory LabCore sample numbers. The subsamples generated in the laboratory for analysis are identified in these diagrams with their sources shown. The diagram identifying the hydrostatic head fluid (HHF) blank is also included, Primary safety screening results and the raw data from Differential Scanning Calorimetry (DSC) and thermogravimetric analysis (TGA) analyses are included in this report. Two of the samples submitted for DSC analysis exceeded notification limits as stated in the Safety Screening DQO (Dukelow, et al., 1995). Cyanide analysis was requested on these samples and a Reactive System Screening Tool analysis was requested for the sample exhibiting the highest exothenn in accordance with the TSAP (Hu, 1996). The results for these analyses will be reported in a revision to this document

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)

Prediction of human core body temperature using non-invasive measurement methods.

Science.gov (United States)

Niedermann, Reto; Wyss, Eva; Annaheim, Simon; Psikuta, Agnes; Davey, Sarah; Rossi, René Michel

2014-01-01

The measurement of core body temperature is an efficient method for monitoring heat stress amongst workers in hot conditions. However, invasive measurement of core body temperature (e.g. rectal, intestinal, oesophageal temperature) is impractical for such applications. Therefore, the aim of this study was to define relevant non-invasive measures to predict core body temperature under various conditions. We conducted two human subject studies with different experimental protocols, different environmental temperatures (10 °C, 30 °C) and different subjects. In both studies the same non-invasive measurement methods (skin temperature, skin heat flux, heart rate) were applied. A principle component analysis was conducted to extract independent factors, which were then used in a linear regression model. We identified six parameters (three skin temperatures, two skin heat fluxes and heart rate), which were included for the calculation of two factors. The predictive value of these factors for core body temperature was evaluated by a multiple regression analysis. The calculated root mean square deviation (rmsd) was in the range from 0.28 °C to 0.34 °C for all environmental conditions. These errors are similar to previous models using non-invasive measures to predict core body temperature. The results from this study illustrate that multiple physiological parameters (e.g. skin temperature and skin heat fluxes) are needed to predict core body temperature. In addition, the physiological measurements chosen in this study and the algorithm defined in this work are potentially applicable as real-time core body temperature monitoring to assess health risk in broad range of working conditions.

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)

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

Identifying Importance-Performance Matrix Analysis (IPMA) of ...

African Journals Online (AJOL)

Identifying Importance-Performance Matrix Analysis (IPMA) of intellectual capital and Islamic work ethics in Malaysian SMES. ... capital and Islamic work ethics significantly influenced business performance. ... AJOL African Journals Online.

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.)

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

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-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999), and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 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 on composite and segment samples; archive half-segment samples; 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-AY-101 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 plans and are not within the scope of this SAP

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

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-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy ''Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO)' (Nguyen 1999a), ''Data Quality Objectives For TWRS Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Butch X (LAW DQO) (Nguyen 1999b)'', ''Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO)'' (Patello et al. 1999), and ''Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO)'' (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 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 on composite and segment samples; archive half-segment samples; and provide sub-samples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AY-101 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 plans and are not within the scope of this SAP

A core-halo pattern of entropy creation in gravitational collapse

Science.gov (United States)

Wren, Andrew J.

2018-03-01

This paper presents a kinetic theory model of gravitational collapse due to a small perturbation. Solving the relevant equations yields a pattern of entropy destruction in a spherical core around the perturbation, and entropy creation in a surrounding halo. This indicates collisional "de-relaxation" in the core, and collisional relaxation in the halo. Core-halo patterns are ubiquitous in the astrophysics of gravitational collapse, and are found here without any of the prior assumptions of such a pattern usually made in analytical models. Motivated by this analysis, the paper outlines a possible scheme for identifying structure formation in a set of observations or a simulation. This scheme involves a choice of coarse-graining scale appropriate to the structure under consideration, and might aid exploration of hierarchical structure formation, supplementing the usual density-based methods for highlighting astrophysical and cosmological structure at various scales.

Event course analysis of core disruptive accidents