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Sample records for analysis identifies core

  1. Identifying functions for ex-core neutron noise analysis

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    Science.gov (United States)

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

    2016-04-01

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

  3. Identifying ELIXIR Core Data Resources.

    Science.gov (United States)

    Durinx, Christine; McEntyre, Jo; Appel, Ron; Apweiler, Rolf; Barlow, Mary; Blomberg, Niklas; Cook, Chuck; Gasteiger, Elisabeth; Kim, Jee-Hyub; Lopez, Rodrigo; Redaschi, Nicole; Stockinger, Heinz; Teixeira, Daniel; Valencia, Alfonso

    2016-01-01

    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.

  4. Comparative Genomic Analysis of Meningitis- and Bacteremia-Causing Pneumococci Identifies a Common Core Genome

    Science.gov (United States)

    Cornick, Jennifer E.; Chaguza, Chrispin; Yalcin, Feyruz; Harris, Simon R.; Gray, Katherine J.; Kiran, Anmol M.; Molyneux, Elizabeth; French, Neil; Faragher, Brian E.; Everett, Dean B.; Bentley, Stephen D.

    2015-01-01

    Streptococcus pneumoniae is a nasopharyngeal commensal that occasionally invades normally sterile sites to cause bloodstream infection and meningitis. Although the pneumococcal population structure and evolutionary genetics are well defined, it is not clear whether pneumococci that cause meningitis are genetically distinct from those that do not. Here, we used whole-genome sequencing of 140 isolates of S. pneumoniae recovered from bloodstream infection (n = 70) and meningitis (n = 70) to compare their genetic contents. By fitting a double-exponential decaying-function model, we show that these isolates share a core of 1,427 genes (95% confidence interval [CI], 1,425 to 1,435 genes) and that there is no difference in the core genome or accessory gene content from these disease manifestations. Gene presence/absence alone therefore does not explain the virulence behavior of pneumococci that reach the meninges. Our analysis, however, supports the requirement of a range of previously described virulence factors and vaccine candidates for both meningitis- and bacteremia-causing pneumococci. This high-resolution view suggests that, despite considerable competency for genetic exchange, all pneumococci are under considerable pressure to retain key components advantageous for colonization and transmission and that these components are essential for access to and survival in sterile sites. PMID:26259813

  5. 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

  6. Identifying regions of strong scattering at the core-mantle boundary from analysis of PKKP precursor energy

    Science.gov (United States)

    Rost, S.; Earle, P.S.

    2010-01-01

    We detect seismic scattering from the core-mantle boundary related to the phase PKKP (PK. KP) in data from small aperture seismic arrays in India and Canada. The detection of these scattered waves in data from small aperture arrays is new and allows a better characterization of the fine-scale structure of the deep Earth especially in the southern hemisphere. Their slowness vector is determined from array processing allowing location of the heterogeneities at the core-mantle boundary using back-projection techniques through 1D Earth models. We identify strong scattering at the core-mantle boundary (CMB) beneath the Caribbean, Patagonia and the Antarctic Peninsula as well as beneath southern Africa. An analysis of the scattering regions relative to sources and receivers indicates that these regions represent areas of increased scattering likely due to increased heterogeneities close to the CMB. The 1. Hz array data used in this study is most sensitive to heterogeneity with scale lengths of about 10. km. Given the small size of the scatterers, a chemical origin of the heterogeneities is likely. By comparing the location of the fine-scale heterogeneity to geodynamical models and tomographic images, we identify different scattering mechanisms in regions related to subduction (Caribbean and Patagonia) and dense thermo chemical piles (Southern Africa). ?? 2010 Elsevier B.V.

  7. 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

  8. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

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

  9. LMFBR core design analysis

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  10. Identifying Core Concepts of Cybersecurity: Results of Two Delphi Processes

    Science.gov (United States)

    Parekh, Geet; DeLatte, David; Herman, Geoffrey L.; Oliva, Linda; Phatak, Dhananjay; Scheponik, Travis; Sherman, Alan T.

    2018-01-01

    This paper presents and analyzes results of two Delphi processes that polled cybersecurity experts to rate cybersecurity topics based on importance, difficulty, and timelessness. These ratings can be used to identify core concepts--cross-cutting ideas that connect knowledge in the discipline. The first Delphi process identified core concepts that…

  11. 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...

  12. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. 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.

  14. 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.

  15. 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.

  16. Identifying the core competencies of mental health telephone triage.

    Science.gov (United States)

    Sands, Natisha; Elsom, Stephen; Gerdtz, Marie; Henderson, Kathryn; Keppich-Arnold, Sandra; Droste, Nicolas; Prematunga, Roshani K; Wereta, Zewdu W

    2013-11-01

    The primary aim of this study was to identify the core competencies of mental health telephone triage, including key role tasks, skills, knowledge and responsibilities, in which clinicians are required to be competent to perform safe and effective triage. Recent global trends indicate an increased reliance on telephone-based health services to facilitate access to health care across large populations. The trend towards telephone-based health services has also extended to mental health settings, evidenced by the growing number of mental health telephone triage services providing 24-hour access to specialist mental health assessment and treatment. Mental health telephone triage services are critical to the early identification of mental health problems and the provision of timely, appropriate interventions. In spite of the rapid growth in mental health telephone triage and the important role these services play in the assessment and management of mental illness and related risks, there has been very little research investigating this area of practice. An observational design was employed to address the research aims. Structured observations (using dual wireless headphones) were undertaken on 197 occasions of mental health telephone triage over a three-month period from January to March 2011. The research identified seven core areas of mental health telephone triage practice in which clinicians are required to be competent in to perform effective mental health telephone triage, including opening the call; performing mental status examination; risk assessment; planning and action; termination of call; referral and reporting; and documentation. The findings of this research contribute to the evidence base for mental health telephone triage by articulating the core competencies for practice. The mental health telephone triage competencies identified in this research may be used to define an evidence-based framework for mental health telephone triage practice that aims to

  17. Core analysis: new features and applications

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Sasaki, L.M.

    1995-01-01

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

  19. Identifying Preserved Storm Events on Beaches from Trenches and Cores

    Science.gov (United States)

    Wadman, H. M.; Gallagher, E. L.; McNinch, J.; Reniers, A.; Koktas, M.

    2014-12-01

    Recent research suggests that even small scale variations in grain size in the shallow stratigraphy of sandy beaches can significantly influence large-scale morphology change. However, few quantitative studies of variations in shallow stratigraphic layers, as differentiated by variations in mean grain size, have been conducted, in no small part due to the difficulty of collecting undisturbed sediment cores in the energetic lower beach and swash zone. Due to this lack of quantitative stratigraphic grain size data, most coastal morphology models assume that uniform grain sizes dominate sandy beaches, allowing for little to no temporal or spatial variations in grain size heterogeneity. In a first-order attempt to quantify small-scale, temporal and spatial variations in beach stratigraphy, thirty-five vibracores were collected at the USACE Field Research Facility (FRF), Duck, NC, in March-April of 2014 using the FRF's Coastal Research and Amphibious Buggy (CRAB). Vibracores were collected at set locations along a cross-shore profile from the toe of the dune to a water depth of ~1m in the surf zone. Vibracores were repeatedly collected from the same locations throughout a tidal cycle, as well as pre- and post a nor'easter event. In addition, two ~1.5m deep trenches were dug in the cross-shore and along-shore directions (each ~14m in length) after coring was completed to allow better interpretation of the stratigraphic sequences observed in the vibracores. The elevations of coherent stratigraphic layers, as revealed in vibracore-based fence diagrams and trench data, are used to relate specific observed stratigraphic sequences to individual storm events observed at the FRF. These data provide a first-order, quantitative examination of the small-scale temporal and spatial variability of shallow grain size along an open, sandy coastline. The data will be used to refine morphological model predictions to include variations in grain size and associated shallow stratigraphy.

  20. 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.

  1. Identifying core domains to assess flare in rheumatoid arthritis

    DEFF Research Database (Denmark)

    Bartlett, Susan J; Hewlett, Sarah; Bingham, Clifton O

    2012-01-01

    For rheumatoid arthritis (RA), there is no consensus on how to define and assess flare. Variability in flare definitions impairs understanding of findings across studies and limits ability to pool results. The OMERACT RA Flare Group sought to identify domains to define RA flares from patient...

  2. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

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

  3. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  4. Identifying the node spreading influence with largest k-core values

    International Nuclear Information System (INIS)

    Lin, Jian-Hong; Guo, Qiang; Dong, Wen-Zhao; Tang, Li-Ying; Liu, Jian-Guo

    2014-01-01

    Identifying the nodes with largest spreading influence of complex networks is one of the most promising domains. By taking into account the neighbors' k-core values, we present an improved neighbors' k-core (INK) method which is the sum of the neighbors' k-core values with a tunable parameter α to evaluate the node spreading influence with largest k-core values. Comparing with the Susceptible–Infected–Recovered (SIR) results for four real networks, the INK method could identify the node spreading influence with largest k-core values more accurately than the ones generated by the degree k, closeness C, betweenness B and coreness centrality method. - Highlights: • We present an improved neighbors' k-core (INK) method to evaluate the node spreading influence with largest k-core values. • The INK method could identify the node spreading influence with largest k-core values more accurately. • Kendall's tau τ of INK method with α=1 are highly identical to rank the node influence

  5. Identifying Core Mobile Learning Faculty Competencies Based Integrated Approach: A Delphi Study

    Science.gov (United States)

    Elbarbary, Rafik Said

    2015-01-01

    This study is based on the integrated approach as a concept framework to identify, categorize, and rank a key component of mobile learning core competencies for Egyptian faculty members in higher education. The field investigation framework used four rounds Delphi technique to determine the importance rate of each component of core competencies…

  6. Statistical hot spot analysis of reactor cores

    International Nuclear Information System (INIS)

    Schaefer, H.

    1974-05-01

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

  7. 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

  8. 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…

  9. A Delphi study to identify the core components of nurse to nurse handoff.

    Science.gov (United States)

    O'Rourke, Jennifer; Abraham, Joanna; Riesenberg, Lee Ann; Matson, Jeff; Lopez, Karen Dunn

    2018-03-08

    The aim of this study was to identify the core components of nurse-nurse handoffs. Patient handoffs involve a process of passing information, responsibility and control from one caregiver to the next during care transitions. Around the globe, ineffective handoffs have serious consequences resulting in wrong treatments, delays in diagnosis, longer stays, medication errors, patient falls and patient deaths. To date, the core components of nurse-nurse handoff have not been identified. This lack of identification is a significant gap in moving towards a standardized approach for nurse-nurse handoff. Mixed methods design using the Delphi technique. From May 2016 - October 2016, using a series of iterative steps, a panel of handoff experts gave feedback on the nurse-nurse handoff core components and the content in each component to be passed from one nurse to the next during a typical unit-based shift handoff. Consensus was defined as 80% agreement or higher. After three rounds of participant review, 17 handoff experts with backgrounds in clinical nursing practice, academia and handoff research came to consensus on the core components of handoff: patient summary, action plan and nurse-nurse synthesis. This is the first study to identify the core components of nurse-nurse handoff. Subsequent testing of the core components will involve evaluating the handoff approach in a simulated and then actual patient care environment. Our long-term goal is to improve patient safety outcomes by validating an evidence-based handoff framework and handoff curriculum for pre-licensure nursing programmes that strengthen the quality of their handoff communication as they enter clinical practice. © 2018 John Wiley & Sons Ltd.

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

    Science.gov (United States)

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

    2006-12-01

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

  11. Identifying Core Competencies of Infection Control Nurse Specialists in Hong Kong.

    Science.gov (United States)

    Chan, Wai Fong; Bond, Trevor G; Adamson, Bob; Chow, Meyrick

    2016-01-01

    To confirm a core competency scale for Hong Kong infection control nurses at the advanced nursing practice level from the core competency items proposed in a previous phase of this study. This would serve as the foundation of competency assurance in Hong Kong hospitals. A cross-sectional survey design was used. All public and private hospitals in Hong Kong. All infection control nurses in hospitals of Hong Kong. The 83-item proposed core competency list established in an earlier study was transformed into a questionnaire and sent to 112 infection control nurses in 48 hospitals in Hong Kong. They were asked to rate the importance of each infection prevention and control item using Likert-style response categories. Data were analyzed using the Rasch model. The response rate of 81.25% was achieved. Seven items were removed from the proposed core competency list, leaving a scale of 76 items that fit the measurement requirements of the unidimensional Rasch model. Essential core competency items of advanced practice for infection control nurses in Hong Kong were identified based on the measurement criteria of the Rasch model. Several items of the scale that reflect local Hong Kong contextual characteristics are distinguished from the overseas standards. This local-specific competency list could serve as the foundation for education and for certification of infection control nurse specialists in Hong Kong. Rasch measurement is an appropriate analytical tool for identifying core competencies of advanced practice nurses in other specialties and in other locations in a manner that incorporates practitioner judgment and expertise.

  12. 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.

  13. 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)

  14. 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)

  15. 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.

  16. 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

  17. 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).

  18. 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)

  19. 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)

  20. 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)

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

    International Nuclear Information System (INIS)

    Bell, K.E.

    1995-01-01

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

  2. 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)

  3. 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.

  4. 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)

  5. Cellular signaling identifiability analysis: a case study.

    Science.gov (United States)

    Roper, Ryan T; Pia Saccomani, Maria; Vicini, Paolo

    2010-05-21

    Two primary purposes for mathematical modeling in cell biology are (1) simulation for making predictions of experimental outcomes and (2) parameter estimation for drawing inferences from experimental data about unobserved aspects of biological systems. While the former purpose has become common in the biological sciences, the latter is less common, particularly when studying cellular and subcellular phenomena such as signaling-the focus of the current study. Data are difficult to obtain at this level. Therefore, even models of only modest complexity can contain parameters for which the available data are insufficient for estimation. In the present study, we use a set of published cellular signaling models to address issues related to global parameter identifiability. That is, we address the following question: assuming known time courses for some model variables, which parameters is it theoretically impossible to estimate, even with continuous, noise-free data? Following an introduction to this problem and its relevance, we perform a full identifiability analysis on a set of cellular signaling models using DAISY (Differential Algebra for the Identifiability of SYstems). We use our analysis to bring to light important issues related to parameter identifiability in ordinary differential equation (ODE) models. We contend that this is, as of yet, an under-appreciated issue in biological modeling and, more particularly, cell biology. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  6. Event course analysis of core disruptive accidents

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  7. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2004-01-01

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

  8. 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.

  9. 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…

  10. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

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

  11. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

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

  12. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2012-01-01

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

  13. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2008-01-01

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

  14. 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)

  15. The Randomised Controlled Trial in Medical Research: Using Bibliometric Methods to Identify Core Journals. A review of: Tsay, Migh-yueh, and Yen-hsu Yang. “Bibliometric Analysis of the Literature of Randomized Controlled Trials.” Journal of the Medical Library Association 93.4 (October 2005: 450-58.

    Directory of Open Access Journals (Sweden)

    John Loy

    2006-03-01

    Full Text Available Objective – To explore the characteristics and distribution of randomized controlled trials (RCTs in the medical literature. The study aims to identify the growth patterns of the RCT, key subject matter, country and language of publication, and determine a list of core journals which contain a substantial proportion of the RCT literature. Design – Retrospective analysis of RCTs. Setting – Medical journal literature. Subjects – A total of 160,213 articles published between 1965‐2001. Detailed analysis of a subset numbering 114,850 articles published from 1990‐2001. Methods – The study seeks to identify all RCTs in MEDLINE from 1965‐2001, and examines the growth rate of the RCT. The authors then do a more detailed analysis on a subset of data from 1990‐2001, using Access database and Excel spreadsheet software, and PERL programming language. The references were analyzed by five fields within MEDLINE; publication type, source, language, country of publication, and descriptor (subject index. Main results – An exponential growth rate for the RCT is demonstrated, suggesting that in the medical literature development has not yet matured and that research using this method continues to grow. A growth rate for the RCT of 11.2% per annum is identified. The most common form of publication is the journal article, making up approximately 98% of the RCT literature. Approximately 75% of the RCTs are multicentre trials indicating that this is the design of choice adopted by researchers. The United States proves to be the greatest source of RCT literature, with 39.9% of journals and 50.6% of articles originating there. After the USA, the most productive countries are England (15.8% of journals and 21.7% articles and Germany (6.5% journals and 6.1% articles. As might be expected, English is the predominant language providing 92.9% of the total publications. Of the remaining 7%, German is the most common language accounting for 2.2%. The top

  16. 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

  17. 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.

  18. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2006-01-01

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

  19. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2005-01-01

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

  20. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2007-01-01

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

  1. Core physics calculation and analysis for SNRE

    International Nuclear Information System (INIS)

    Xie Jiachun; Zhao Shouzhi; Jia Baoshan

    2010-01-01

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

  2. 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)

  3. 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.

  4. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

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

    1971-01-01

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

  5. Core management and performance analysis for PWR

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  6. 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

  7. 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).

  8. 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.)

  9. 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

  10. A core ontology for business process analysis

    NARCIS (Netherlands)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. 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

  13. 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)

  14. 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

  15. 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)

  16. 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 ...

  17. 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.

  18. Identifying Pornographic Materials with Judgment Analysis

    Science.gov (United States)

    Houston, Judith A.; Houston, Samuel R.

    1974-01-01

    The primary purpose of this study was to determine if a policy-capturing methodology (JAN) which has been successfully utilized in military and educational research could be adapted for use as a procedure in identifying pornographic material. (Author)

  19. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  20. 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.

  1. 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.

  2. 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.

  3. Identifiable Data Files - Medicare Provider Analysis and ...

    Data.gov (United States)

    U.S. Department of Health & Human Services — The Medicare Provider Analysis and Review (MEDPAR) File contains data from claims for services provided to beneficiaries admitted to Medicare certified inpatient...

  4. 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.

  5. Identifying MMORPG Bots: A Traffic Analysis Approach

    Science.gov (United States)

    Chen, Kuan-Ta; Jiang, Jhih-Wei; Huang, Polly; Chu, Hao-Hua; Lei, Chin-Laung; Chen, Wen-Chin

    2008-12-01

    Massively multiplayer online role playing games (MMORPGs) have become extremely popular among network gamers. Despite their success, one of MMORPG's greatest challenges is the increasing use of game bots, that is, autoplaying game clients. The use of game bots is considered unsportsmanlike and is therefore forbidden. To keep games in order, game police, played by actual human players, often patrol game zones and question suspicious players. This practice, however, is labor-intensive and ineffective. To address this problem, we analyze the traffic generated by human players versus game bots and propose general solutions to identify game bots. Taking Ragnarok Online as our subject, we study the traffic generated by human players and game bots. We find that their traffic is distinguishable by 1) the regularity in the release time of client commands, 2) the trend and magnitude of traffic burstiness in multiple time scales, and 3) the sensitivity to different network conditions. Based on these findings, we propose four strategies and two ensemble schemes to identify bots. Finally, we discuss the robustness of the proposed methods against countermeasures of bot developers, and consider a number of possible ways to manage the increasingly serious bot problem.

  6. Identifying MMORPG Bots: A Traffic Analysis Approach

    Directory of Open Access Journals (Sweden)

    Wen-Chin Chen

    2008-11-01

    Full Text Available Massively multiplayer online role playing games (MMORPGs have become extremely popular among network gamers. Despite their success, one of MMORPG's greatest challenges is the increasing use of game bots, that is, autoplaying game clients. The use of game bots is considered unsportsmanlike and is therefore forbidden. To keep games in order, game police, played by actual human players, often patrol game zones and question suspicious players. This practice, however, is labor-intensive and ineffective. To address this problem, we analyze the traffic generated by human players versus game bots and propose general solutions to identify game bots. Taking Ragnarok Online as our subject, we study the traffic generated by human players and game bots. We find that their traffic is distinguishable by 1 the regularity in the release time of client commands, 2 the trend and magnitude of traffic burstiness in multiple time scales, and 3 the sensitivity to different network conditions. Based on these findings, we propose four strategies and two ensemble schemes to identify bots. Finally, we discuss the robustness of the proposed methods against countermeasures of bot developers, and consider a number of possible ways to manage the increasingly serious bot problem.

  7. 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

  8. 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 ...

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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)

  19. Identifying core, exciting, and hybrid attributes in fans' assessments of major (World Cup) spectator sports events.

    Science.gov (United States)

    Gau, Li-Shiue

    2013-12-01

    This paper adopts a methodology of asymmetrical analyses to investigate the relevant importance of spectator sport attributes in terms of their non-linear associations with the benefits that fans experience while watching sports. Questionnaires tapping 16 attributes (e.g., teamwork, sportsmanship, level of competition) and 16 benefits (e.g., good mood, exciting experience, support for my favorite team) were distributed to a sample of fans at the outdoor broadcast of the 2010 World Cup final game at the National Stadium in Kaohsiung, Taiwan. 427 participants rated the importance and benefit of each attribute experienced from watching the game. Attributes were categorized as core, exciting, or hybrid attributes. The star player was the core attribute; rivalry, popularity, and coach were the exciting attributes; and the other 12 attributes were hybrid. Two-dimensional space analyses showed that attributes "sportsmanship, teamwork, and supporting a team" were both explicitly and implicitly important attributes. The methodology of asymmetrical analyses can help managers prioritize the focus of attributes and allocate resources effectively.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

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

    1990-04-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. Identifying the core microbial community in the gut of fungus-growing termites

    DEFF Research Database (Denmark)

    Otani, Saria; Mikaelyan, Aram; Nobre, Tânia

    2014-01-01

    Gut microbes play a crucial role in decomposing lignocellulose to fuel termite societies, with protists in the lower termites and prokaryotes in the higher termites providing these services. However, a single basal subfamily of the higher termites, the Macrotermitinae, also domesticated a plant......, and Synergistetes. A set of 42 genus-level taxa was present in all termite species and accounted for 56-68% of the species-specific reads. Gut communities of termites from the same genus were more similar than distantly related species, suggesting that phylogenetic ancestry matters, possibly in connection...... with specific termite genus-level ecological niches. Finally, we show that gut communities of fungus-growing termites are similar to cockroaches, both at the bacterial phylum level and in a comparison of the core Macrotermitinae taxa abundances with representative cockroach, lower termite, and higher non...

  4. 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

  5. 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

  6. 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.

  7. 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.

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

    Science.gov (United States)

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

    2016-09-23

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

  9. 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.)

  10. 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.

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. Microbial Analysis of Australian Dry Lake Cores; Analogs For Biogeochemical Processes

    Science.gov (United States)

    Nguyen, A. V.; Baldridge, A. M.; Thomson, B. J.

    2014-12-01

    Lake Gilmore in Western Australia is an acidic ephemeral lake that is analogous to Martian geochemical processes represented by interbedded phyllosilicates and sulfates. These areas demonstrate remnants of a global-scale change on Mars during the late Noachian era from a neutral to alkaline pH to relatively lower pH in the Hesperian era that continues to persist today. The geochemistry of these areas could possibly be caused by small-scale changes such as microbial metabolism. Two approaches were used to determine the presence of microbes in the Australian dry lake cores: DNA analysis and lipid analysis. Detecting DNA or lipids in the cores will provide evidence of living or deceased organisms since they provide distinct markers for life. Basic DNA analysis consists of extraction, amplification through PCR, plasmid cloning, and DNA sequencing. Once the sequence of unknown DNA is known, an online program, BLAST, will be used to identify the microbes for further analysis. The lipid analysis approach consists of phospholipid fatty acid analysis that is done by Microbial ID, which will provide direct identification any microbes from the presence of lipids. Identified microbes are then compared to mineralogy results from the x-ray diffraction of the core samples to determine if the types of metabolic reactions are consistent with the variation in composition in these analog deposits. If so, it provides intriguing implications for the presence of life in similar Martian deposits.

  1. Postsecondary Students With Psychiatric Disabilities Identify Core Services and Key Ingredients to Supporting Education Goals.

    Science.gov (United States)

    Biebel, Kathleen; Mizrahi, Raphael; Ringeisen, Heather

    2017-10-26

    Accessing and successfully completing postsecondary educational opportunities may be challenging for those living with psychiatric disabilities. This exploratory study highlights the experiences of individuals with psychiatric disabilities participating in postsecondary educational support initiatives. Investigators conducted case studies with 3 education support initiatives across the United States. Focus groups revealed what concrete supported education services were helpful and key ingredients in delivering education supports. Access to specialists, mindfulness techniques, help with time management and procrastination, and facilitating classroom accommodations were identified as critical. Developing authentic relationships with supported education staff, flexibility in service delivery and access to student peers living with psychiatric disabilities were noted as key ingredients in service delivery. Incorporating the voice of students with psychiatric disabilities into supported education services can increase access, involvement, and retention, therein providing more supports to students with psychiatric disabilities achieving their postsecondary education goals. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  2. 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

  3. 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

  4. Minimally Actuated Walking: Identifying Core Challenges to Economical Legged Locomotion Reveals Novel Solutions

    Directory of Open Access Journals (Sweden)

    Ryan T Schroeder

    2018-05-01

    Full Text Available Terrestrial organisms adept at locomotion employ strut-like legs for economical and robust movement across the substrate. Although it is relatively easy to observe and analyze details of the solutions these organic systems have arrived at, it is not as easy to identify the problems these movement strategies have solved. As such, it is useful to investigate fundamental challenges that effective legged locomotion overcomes in order to understand why the mechanisms employed by biological systems provide viable solutions to these challenges. Such insight can inform the design and development of legged robots that may eventually match or exceed animal performance. In the context of human walking, we apply control optimization as a design strategy for simple bipedal walking machines with minimal actuation. This approach is used to discuss key facilitators of energetically efficient locomotion in simple bipedal walkers. Furthermore, we extrapolate the approach to a novel application—a theoretical exoskeleton attached to the trunk of a human walker—to demonstrate how coordinated efforts between bipedal actuation and a machine oscillator can potentially alleviate a meaningful portion of energetic exertion associated with leg function during human walking.

  5. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  7. Case for integral core-disruptive accident analysis

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  8. 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

  9. 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.

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Fan Kai; Li Fu; Zhou Xuhua

    2014-01-01

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

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

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

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

  17. 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)

  18. Performance Analysis: Work Control Events Identified January - August 2010

    Energy Technology Data Exchange (ETDEWEB)

    De Grange, C E; Freeman, J W; Kerr, C E; Holman, G; Marsh, K; Beach, R

    2011-01-14

    This performance analysis evaluated 24 events that occurred at LLNL from January through August 2010. The analysis identified areas of potential work control process and/or implementation weaknesses and several common underlying causes. Human performance improvement and safety culture factors were part of the causal analysis of each event and were analyzed. The collective significance of all events in 2010, as measured by the occurrence reporting significance category and by the proportion of events that have been reported to the DOE ORPS under the ''management concerns'' reporting criteria, does not appear to have increased in 2010. The frequency of reporting in each of the significance categories has not changed in 2010 compared to the previous four years. There is no change indicating a trend in the significance category and there has been no increase in the proportion of occurrences reported in the higher significance category. Also, the frequency of events, 42 events reported through August 2010, is not greater than in previous years and is below the average of 63 occurrences per year at LLNL since 2006. Over the previous four years, an average of 43% of the LLNL's reported occurrences have been reported as either ''management concerns'' or ''near misses.'' In 2010, 29% of the occurrences have been reported as ''management concerns'' or ''near misses.'' This rate indicates that LLNL is now reporting fewer ''management concern'' and ''near miss'' occurrences compared to the previous four years. From 2008 to the present, LLNL senior management has undertaken a series of initiatives to strengthen the work planning and control system with the primary objective to improve worker safety. In 2008, the LLNL Deputy Director established the Work Control Integrated Project Team to develop the core requirements and graded

  19. 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.

  20. NREL Analysis Identifies Where Commercial Customers Might Benefit from

    Science.gov (United States)

    Battery Energy Storage | NREL | News | NREL NREL Analysis Identifies Where Commercial Customers Customers Might Benefit from Battery Energy Storage August 24, 2017 After upfront costs, batteries may reduce operating costs for customers paying demand charges Commercial electricity customers who are

  1. 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.

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

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

    1998-03-01

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

  5. 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)

  6. 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)

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

    DEFF Research Database (Denmark)

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

    2016-01-01

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

  8. 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)

  9. 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

  10. 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.

  11. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. Identifying Core Herbal Treatments for Children with Asthma: Implication from a Chinese Herbal Medicine Database in Taiwan

    Directory of Open Access Journals (Sweden)

    Hsing-Yu Chen

    2013-01-01

    Full Text Available Asthma is one of the most common allergic respiratory diseases around the world and places great burden on medical payment. Chinese herbal medicine (CHM is commonly used for Taiwanese children to control diseases. The aim of this study is to analyze the CHM prescriptions for asthmatic children by using a nationwide clinical database. The National Health Insurance Research Database (NHIRD was used to perform this study. Medical records from 1997 to 2009 with diagnosis with asthma made for children aged 6 to 18 were included into the analysis. Association rule mining and social network analysis were used to analyze the prevalence of single CHM and its combinations. Ma-Xing-Gan-Shi-Tang (MXGST was the most commonly used herbal formula (HF (20.2% of all prescriptions, followed by Xiao-Qing-Long-Tang (13.1% and Xing-Su-San (12.8%. Zhe Bei Mu is the most frequently used single herb (SH (14.6%, followed by Xing Ren (10.7%. MXGST was commonly used with Zhe Bei Mu (3.5% and other single herbs capable of dispelling phlegm. Besides, MXGST was the core formula to relieve asthma. Further studies about efficacy and drug safety are needed for the CHM commonly used for asthma based on the result of this study.

  2. Identifying core herbal treatments for children with asthma: implication from a chinese herbal medicine database in taiwan.

    Science.gov (United States)

    Chen, Hsing-Yu; Lin, Yi-Hsuan; Thien, Peck-Foong; Chang, Shih-Chieh; Chen, Yu-Chun; Lo, Su-Shun; Yang, Sien-Hung; Chen, Jiun-Liang

    2013-01-01

    Asthma is one of the most common allergic respiratory diseases around the world and places great burden on medical payment. Chinese herbal medicine (CHM) is commonly used for Taiwanese children to control diseases. The aim of this study is to analyze the CHM prescriptions for asthmatic children by using a nationwide clinical database. The National Health Insurance Research Database (NHIRD) was used to perform this study. Medical records from 1997 to 2009 with diagnosis with asthma made for children aged 6 to 18 were included into the analysis. Association rule mining and social network analysis were used to analyze the prevalence of single CHM and its combinations. Ma-Xing-Gan-Shi-Tang (MXGST) was the most commonly used herbal formula (HF) (20.2% of all prescriptions), followed by Xiao-Qing-Long-Tang (13.1%) and Xing-Su-San (12.8%). Zhe Bei Mu is the most frequently used single herb (SH) (14.6%), followed by Xing Ren (10.7%). MXGST was commonly used with Zhe Bei Mu (3.5%) and other single herbs capable of dispelling phlegm. Besides, MXGST was the core formula to relieve asthma. Further studies about efficacy and drug safety are needed for the CHM commonly used for asthma based on the result of this study.

  3. Development of advanced nuclear core analysis system applicable to various reactor types

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

  4. 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

  5. BIOELECTRICAL IMPEDANCE VECTOR ANALYSIS IDENTIFIES SARCOPENIA IN NURSING HOME RESIDENTS

    Science.gov (United States)

    Loss of muscle mass and water shifts between body compartments are contributing factors to frailty in the elderly. The body composition changes are especially pronounced in institutionalized elderly. We investigated the ability of single-frequency bioelectrical impedance analysis (BIA) to identify b...

  6. Identifying Students’ Misconceptions on Basic Algorithmic Concepts Through Flowchart Analysis

    NARCIS (Netherlands)

    Rahimi, E.; Barendsen, E.; Henze, I.; Dagienė, V.; Hellas, A.

    2017-01-01

    In this paper, a flowchart-based approach to identifying secondary school students’ misconceptions (in a broad sense) on basic algorithm concepts is introduced. This approach uses student-generated flowcharts as the units of analysis and examines them against plan composition and construct-based

  7. 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

  8. Heating analysis of cobalt adjusters in reactor core

    International Nuclear Information System (INIS)

    Mei Qiliang; Li Kang; Fu Yaru

    2011-01-01

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

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

    KAUST Repository

    Schrittwieser, Stefan

    2016-03-29

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

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

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

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

  11. 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.

  12. Structural parameter identifiability analysis for dynamic reaction networks

    DEFF Research Database (Denmark)

    Davidescu, Florin Paul; Jørgensen, Sten Bay

    2008-01-01

    method based on Lie derivatives. The proposed systematic two phase methodology is illustrated on a mass action based model for an enzymatically catalyzed reaction pathway network where only a limited set of variables is measured. The methodology clearly pinpoints the structurally identifiable parameters...... where for a given set of measured variables it is desirable to investigate which parameters may be estimated prior to spending computational effort on the actual estimation. This contribution addresses the structural parameter identifiability problem for the typical case of reaction network models....... The proposed analysis is performed in two phases. The first phase determines the structurally identifiable reaction rates based on reaction network stoichiometry. The second phase assesses the structural parameter identifiability of the specific kinetic rate expressions using a generating series expansion...

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

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

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

  14. 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

  15. 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

  16. 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

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

    African Journals Online (AJOL)

    2016-07-27

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

  18. Identifying clinical course patterns in SMS data using cluster analysis

    DEFF Research Database (Denmark)

    Kent, Peter; Kongsted, Alice

    2012-01-01

    ABSTRACT: BACKGROUND: Recently, there has been interest in using the short message service (SMS or text messaging), to gather frequent information on the clinical course of individual patients. One possible role for identifying clinical course patterns is to assist in exploring clinically important...... showed that clinical course patterns can be identified by cluster analysis using all SMS time points as cluster variables. This method is simple, intuitive and does not require a high level of statistical skill. However, there are alternative ways of managing SMS data and many different methods...

  19. 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

  20. 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.

  1. 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.

  2. 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)

  3. 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

  4. Latent cluster analysis of ALS phenotypes identifies prognostically differing groups.

    Directory of Open Access Journals (Sweden)

    Jeban Ganesalingam

    2009-09-01

    Full Text Available Amyotrophic lateral sclerosis (ALS is a degenerative disease predominantly affecting motor neurons and manifesting as several different phenotypes. Whether these phenotypes correspond to different underlying disease processes is unknown. We used latent cluster analysis to identify groupings of clinical variables in an objective and unbiased way to improve phenotyping for clinical and research purposes.Latent class cluster analysis was applied to a large database consisting of 1467 records of people with ALS, using discrete variables which can be readily determined at the first clinic appointment. The model was tested for clinical relevance by survival analysis of the phenotypic groupings using the Kaplan-Meier method.The best model generated five distinct phenotypic classes that strongly predicted survival (p<0.0001. Eight variables were used for the latent class analysis, but a good estimate of the classification could be obtained using just two variables: site of first symptoms (bulbar or limb and time from symptom onset to diagnosis (p<0.00001.The five phenotypic classes identified using latent cluster analysis can predict prognosis. They could be used to stratify patients recruited into clinical trials and generating more homogeneous disease groups for genetic, proteomic and risk factor research.

  5. Use of Photogrammetry and Biomechanical Gait analysis to Identify Individuals

    DEFF Research Database (Denmark)

    Larsen, Peter Kastmand; Simonsen, Erik Bruun; Lynnerup, Niels

    Photogrammetry and recognition of gait patterns are valuable tools to help identify perpetrators based on surveillance recordings. We have found that stature but only few other measures have a satisfying reproducibility for use in forensics. Several gait variables with high recognition rates were...... found. Especially the variables located in the frontal plane are interesting due to large inter-individual differences in time course patterns. The variables with high recognition rates seem preferable for use in forensic gait analysis and as input variables to waveform analysis techniques...

  6. Identifying influential factors of business process performance using dependency analysis

    Science.gov (United States)

    Wetzstein, Branimir; Leitner, Philipp; Rosenberg, Florian; Dustdar, Schahram; Leymann, Frank

    2011-02-01

    We present a comprehensive framework for identifying influential factors of business process performance. In particular, our approach combines monitoring of process events and Quality of Service (QoS) measurements with dependency analysis to effectively identify influential factors. The framework uses data mining techniques to construct tree structures to represent dependencies of a key performance indicator (KPI) on process and QoS metrics. These dependency trees allow business analysts to determine how process KPIs depend on lower-level process metrics and QoS characteristics of the IT infrastructure. The structure of the dependencies enables a drill-down analysis of single factors of influence to gain a deeper knowledge why certain KPI targets are not met.

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

    Science.gov (United States)

    Zheng, Siwen; Liu, Yazhuo; Chang, Guangjian

    2018-03-01

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

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

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

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

  9. 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)

  10. 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

  11. Identifying Organizational Inefficiencies with Pictorial Process Analysis (PPA

    Directory of Open Access Journals (Sweden)

    David John Patrishkoff

    2013-11-01

    Full Text Available Pictorial Process Analysis (PPA was created by the author in 2004. PPA is a unique methodology which offers ten layers of additional analysis when compared to standard process mapping techniques.  The goal of PPA is to identify and eliminate waste, inefficiencies and risk in manufacturing or transactional business processes at 5 levels in an organization. The highest level being assessed is the process management, followed by the process work environment, detailed work habits, process performance metrics and general attitudes towards the process. This detailed process assessment and analysis is carried out during process improvement brainstorming efforts and Kaizen events. PPA creates a detailed visual efficiency rating for each step of the process under review.  A selection of 54 pictorial Inefficiency Icons (cards are available for use to highlight major inefficiencies and risks that are present in the business process under review. These inefficiency icons were identified during the author's independent research on the topic of why things go wrong in business. This paper will highlight how PPA was developed and show the steps required to conduct Pictorial Process Analysis on a sample manufacturing process. The author has successfully used PPA to dramatically improve business processes in over 55 different industries since 2004.  

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

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

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

  13. 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.

  14. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Zhang Yabo; Yang Hongyi

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. 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)

  19. 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

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

    International Nuclear Information System (INIS)

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

    1996-02-01

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

  1. 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.

  2. Rice Transcriptome Analysis to Identify Possible Herbicide Quinclorac Detoxification Genes

    Directory of Open Access Journals (Sweden)

    Wenying eXu

    2015-09-01

    Full Text Available Quinclorac is a highly selective auxin-type herbicide, and is widely used in the effective control of barnyard grass in paddy rice fields, improving the world’s rice yield. The herbicide mode of action of quinclorac has been proposed and hormone interactions affect quinclorac signaling. Because of widespread use, quinclorac may be transported outside rice fields with the drainage waters, leading to soil and water pollution and environmental health problems.In this study, we used 57K Affymetrix rice whole-genome array to identify quinclorac signaling response genes to study the molecular mechanisms of action and detoxification of quinclorac in rice plants. Overall, 637 probe sets were identified with differential expression levels under either 6 or 24 h of quinclorac treatment. Auxin-related genes such as GH3 and OsIAAs responded to quinclorac treatment. Gene Ontology analysis showed that genes of detoxification-related family genes were significantly enriched, including cytochrome P450, GST, UGT, and ABC and drug transporter genes. Moreover, real-time RT-PCR analysis showed that top candidate P450 families such as CYP81, CYP709C and CYP72A genes were universally induced by different herbicides. Some Arabidopsis genes for the same P450 family were up-regulated under quinclorac treatment.We conduct rice whole-genome GeneChip analysis and the first global identification of quinclorac response genes. This work may provide potential markers for detoxification of quinclorac and biomonitors of environmental chemical pollution.

  3. 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.

  4. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Science.gov (United States)

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

    1977-01-01

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

  7. Obesogenic family types identified through latent profile analysis.

    Science.gov (United States)

    Martinson, Brian C; VazquezBenitez, Gabriela; Patnode, Carrie D; Hearst, Mary O; Sherwood, Nancy E; Parker, Emily D; Sirard, John; Pasch, Keryn E; Lytle, Leslie

    2011-10-01

    Obesity may cluster in families due to shared physical and social environments. This study aims to identify family typologies of obesity risk based on family environments. Using 2007-2008 data from 706 parent/youth dyads in Minnesota, we applied latent profile analysis and general linear models to evaluate associations between family typologies and body mass index (BMI) of youth and parents. Three typologies described most families with 18.8% "Unenriched/Obesogenic," 16.9% "Risky Consumer," and 64.3% "Healthy Consumer/Salutogenic." After adjustment for demographic and socioeconomic factors, parent BMI and youth BMI Z-scores were higher in unenriched/obesogenic families (BMI difference = 2.7, p typology. In contrast, parent BMI and youth BMI Z-scores were similar in the risky consumer families relative to those in healthy consumer/salutogenic type. We can identify family types differing in obesity risks with implications for public health interventions.

  8. Proteogenomic Analysis Identifies a Novel Human SHANK3 Isoform

    Directory of Open Access Journals (Sweden)

    Fahad Benthani

    2015-05-01

    Full Text Available Mutations of the SHANK3 gene have been associated with autism spectrum disorder. Individuals harboring different SHANK3 mutations display considerable heterogeneity in their cognitive impairment, likely due to the high SHANK3 transcriptional diversity. In this study, we report a novel interaction between the Mutated in colorectal cancer (MCC protein and a newly identified SHANK3 protein isoform in human colon cancer cells and mouse brain tissue. Hence, our proteogenomic analysis identifies a new human long isoform of the key synaptic protein SHANK3 that was not predicted by the human reference genome. Taken together, our findings describe a potential new role for MCC in neurons, a new human SHANK3 long isoform and, importantly, highlight the use of proteomic data towards the re-annotation of GC-rich genomic regions.

  9. 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

  10. Parameter trajectory analysis to identify treatment effects of pharmacological interventions.

    Directory of Open Access Journals (Sweden)

    Christian A Tiemann

    Full Text Available The field of medical systems biology aims to advance understanding of molecular mechanisms that drive disease progression and to translate this knowledge into therapies to effectively treat diseases. A challenging task is the investigation of long-term effects of a (pharmacological treatment, to establish its applicability and to identify potential side effects. We present a new modeling approach, called Analysis of Dynamic Adaptations in Parameter Trajectories (ADAPT, to analyze the long-term effects of a pharmacological intervention. A concept of time-dependent evolution of model parameters is introduced to study the dynamics of molecular adaptations. The progression of these adaptations is predicted by identifying necessary dynamic changes in the model parameters to describe the transition between experimental data obtained during different stages of the treatment. The trajectories provide insight in the affected underlying biological systems and identify the molecular events that should be studied in more detail to unravel the mechanistic basis of treatment outcome. Modulating effects caused by interactions with the proteome and transcriptome levels, which are often less well understood, can be captured by the time-dependent descriptions of the parameters. ADAPT was employed to identify metabolic adaptations induced upon pharmacological activation of the liver X receptor (LXR, a potential drug target to treat or prevent atherosclerosis. The trajectories were investigated to study the cascade of adaptations. This provided a counter-intuitive insight concerning the function of scavenger receptor class B1 (SR-B1, a receptor that facilitates the hepatic uptake of cholesterol. Although activation of LXR promotes cholesterol efflux and -excretion, our computational analysis showed that the hepatic capacity to clear cholesterol was reduced upon prolonged treatment. This prediction was confirmed experimentally by immunoblotting measurements of SR-B1

  11. Identifying inaccuracy of MS Project using system analysis

    Science.gov (United States)

    Fachrurrazi; Husin, Saiful; Malahayati, Nurul; Irzaidi

    2018-05-01

    The problem encountered in project owner’s financial accounting report is the difference in total project costs of MS Project to the Indonesian Standard (Standard Indonesia Standard / Cost Estimating Standard Book of Indonesia). It is one of the MS Project problems concerning to its cost accuracy, so cost data cannot be used in an integrated way for all project components. This study focuses on finding the causes of inaccuracy of the MS Projects. The aim of this study, which is operationally, are: (i) identifying cost analysis procedures for both current methods (SNI) and MS Project; (ii) identifying cost bias in each element of the cost analysis procedure; and (iii) analysing the cost differences (cost bias) in each element to identify what the cause of inaccuracies in MS Project toward SNI is. The method in this study is comparing for both the system analysis of MS Project and SNI. The results are: (i) MS Project system in Work of Resources element has limitation for two decimal digits only, have led to its inaccuracy. Where the Work of Resources (referred to as effort) in MS Project represents multiplication between the Quantities of Activities and Requirements of resources in SNI; (ii) MS Project and SNI have differences in the costing methods (the cost estimation methods), in which the SNI uses the Quantity-Based Costing (QBC), meanwhile MS Project uses the Time-Based Costing (TBC). Based on this research, we recommend to the contractors who use SNI should make an adjustment for Work of Resources in MS Project (with correction index) so that it can be used in an integrated way to the project owner’s financial accounting system. Further research will conduct for improvement the MS Project as an integrated tool toward all part of the project participant.

  12. Core subjects at the end of primary school: identifying and explaining relative strengths of children with specific language impairment (SLI)

    Science.gov (United States)

    Durkin, Kevin; Mok, Pearl L H; Conti-Ramsden, Gina

    2015-01-01

    Background In general, children with specific language impairment (SLI) tend to fall behind their typically developing (TD) peers in educational attainment. Less is known about how children with SLI fare in particular areas of the curriculum and what predicts their levels of performance. Aims To compare the distributions of performance of children with SLI in three core school subjects (English, Mathematics and Science); to test the possibility that performance would vary across the core subjects; and to examine the extent to which language impairment predicts performance. Methods & Procedures This study was conducted in England and reports historical data on educational attainments. Teacher assessment and test scores of 176 eleven-year-old children with SLI were examined in the three core subjects and compared with known national norms. Possible predictors of performance were measured, including language ability at ages 7 and 11, educational placement type, and performance IQ. Outcomes & Results Children with SLI, compared with national norms, were found to be at a disadvantage in core school subjects. Nevertheless, some children attained the levels expected of TD peers. Performance was poorest in English; relative strengths were indicated in Science and, to a lesser extent, in Mathematics. Language skills were significant predictors of performance in all three core subjects. PIQ was the strongest predictor for Mathematics. For Science, both early language skills at 7 years and PIQ made significant contributions. Conclusions & Implications Language impacts on the school performance of children with SLI, but differentially across subjects. English for these children is the most challenging of the core subjects, reflecting the high levels of language demand it incurs. Science is an area of relative strength and mathematics appears to be intermediate, arguably because some tasks in these subjects can be performed with less reliance on verbal processing. Many children

  13. 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.)

  14. 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)

  15. 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.

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

    International Nuclear Information System (INIS)

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

    2002-04-01

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

  17. Using lexical analysis to identify emotional distress in psychometric schizotypy.

    Science.gov (United States)

    Abplanalp, Samuel J; Buck, Benjamin; Gonzenbach, Virgilio; Janela, Carlos; Lysaker, Paul H; Minor, Kyle S

    2017-09-01

    Through the use of lexical analysis software, researchers have demonstrated a greater frequency of negative affect word use in those with schizophrenia and schizotypy compared to the general population. In addition, those with schizotypy endorse greater emotional distress than healthy controls. In this study, our aim was to expand on previous findings in schizotypy to determine whether negative affect word use could be linked to emotional distress. Schizotypy (n=33) and non-schizotypy groups (n=33) completed an open-ended, semi-structured interview and negative affect word use was analyzed using a validated lexical analysis instrument. Emotional distress was assessed using subjective questionnaires of depression and psychological quality of life (QOL). When groups were compared, those with schizotypy used significantly more negative affect words; endorsed greater depression; and reported lower QOL. Within schizotypy, a trend level association between depression and negative affect word use was observed; QOL and negative affect word use showed a significant inverse association. Our findings offer preliminary evidence of the potential effectiveness of lexical analysis as an objective, behavior-based method for identifying emotional distress throughout the schizophrenia-spectrum. Utilizing lexical analysis in schizotypy offers promise for providing researchers with an assessment capable of objectively detecting emotional distress. Copyright © 2017 Elsevier Ireland Ltd. All rights reserved.

  18. Cytokeratin 5 and estrogen receptor immunohistochemistry as a useful adjunct in identifying atypical papillary lesions on breast needle core biopsy.

    Science.gov (United States)

    Grin, Andrea; O'Malley, Frances P; Mulligan, Anna Marie

    2009-11-01

    The presence of atypical or usual epithelial proliferations within papillary breast lesions complicates their interpretation on core biopsy. We evaluated the combination of estrogen receptor (ER) and cytokeratin 5 (CK5) as an aid in the distinction of usual duct hyperplasia from atypical proliferations in this setting. Core biopsies from 185 papillary lesions were reviewed and of these, 82 cases were selected for immunohistochemical study based on the presence of an epithelial proliferation between the fibrovascular cores. Fifty-two cases were used as the test set and 30 cases, with subsequent surgical excision, were used as the validation set. The epithelial proliferation was evaluated for staining intensity and percentage of positive cells using CK5 and ER. Expression of both CK5 and ER was significantly different in nonatypical lesions when compared with atypical lesions (P90% of cells. CK5-high expression was defined as a mosaic pattern of staining in >20% of cells and CK5-low as absent or staining in hyperplasia from atypical proliferations within papillary lesions on core biopsy.

  19. Using imagery to identify and characterise core beliefs in women with bulimia nervosa, dieting and non-dieting women.

    Science.gov (United States)

    Somerville, Kate; Cooper, Myra

    2007-12-01

    Women with bulimia nervosa (BN), dieters and non-dieting control participants were questioned about spontaneous imagery linked to concern with food and eating, weight and shape. The downward arrow technique was used to access any associated negative or core beliefs, which were examined for belief, distress and content. A semi-structured interview with open and closed questions was used. Negative self (core) beliefs were successfully accessed, and responses to the interview items had good test-retest and good inter-rater reliability. Patients with BN reported significantly more negative self (core) beliefs than those in the other two groups. Only a very small number of core beliefs about other people or the world in general were reported. Emotional belief ratings appeared to be higher overall than rational belief ratings. Patient's negative self-beliefs contained themes of "self-value", followed by "failure", "self-control" and "physical attractiveness", in descending order of frequency. The findings are discussed in relation to existing research, and implications for cognitive theories of bulimia nervosa and clinical practice are briefly discussed.

  20. Cluster analysis of clinical data identifies fibromyalgia subgroups.

    Directory of Open Access Journals (Sweden)

    Elisa Docampo

    Full Text Available INTRODUCTION: Fibromyalgia (FM is mainly characterized by widespread pain and multiple accompanying symptoms, which hinder FM assessment and management. In order to reduce FM heterogeneity we classified clinical data into simplified dimensions that were used to define FM subgroups. MATERIAL AND METHODS: 48 variables were evaluated in 1,446 Spanish FM cases fulfilling 1990 ACR FM criteria. A partitioning analysis was performed to find groups of variables similar to each other. Similarities between variables were identified and the variables were grouped into dimensions. This was performed in a subset of 559 patients, and cross-validated in the remaining 887 patients. For each sample and dimension, a composite index was obtained based on the weights of the variables included in the dimension. Finally, a clustering procedure was applied to the indexes, resulting in FM subgroups. RESULTS: VARIABLES CLUSTERED INTO THREE INDEPENDENT DIMENSIONS: "symptomatology", "comorbidities" and "clinical scales". Only the two first dimensions were considered for the construction of FM subgroups. Resulting scores classified FM samples into three subgroups: low symptomatology and comorbidities (Cluster 1, high symptomatology and comorbidities (Cluster 2, and high symptomatology but low comorbidities (Cluster 3, showing differences in measures of disease severity. CONCLUSIONS: We have identified three subgroups of FM samples in a large cohort of FM by clustering clinical data. Our analysis stresses the importance of family and personal history of FM comorbidities. Also, the resulting patient clusters could indicate different forms of the disease, relevant to future research, and might have an impact on clinical assessment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  2. Analysis of core samples from jet grouted soil

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  3. 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)

  4. 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

  5. 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)

  6. 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...

  7. Core subjects at the end of primary school: identifying and explaining relative strengths of children with specific language impairment (SLI).

    Science.gov (United States)

    Durkin, Kevin; Mok, Pearl L H; Conti-Ramsden, Gina

    2015-01-01

    In general, children with specific language impairment (SLI) tend to fall behind their typically developing (TD) peers in educational attainment. Less is known about how children with SLI fare in particular areas of the curriculum and what predicts their levels of performance. To compare the distributions of performance of children with SLI in three core school subjects (English, Mathematics and Science); to test the possibility that performance would vary across the core subjects; and to examine the extent to which language impairment predicts performance. This study was conducted in England and reports historical data on educational attainments. Teacher assessment and test scores of 176 eleven-year-old children with SLI were examined in the three core subjects and compared with known national norms. Possible predictors of performance were measured, including language ability at ages 7 and 11, educational placement type, and performance IQ. Children with SLI, compared with national norms, were found to be at a disadvantage in core school subjects. Nevertheless, some children attained the levels expected of TD peers. Performance was poorest in English; relative strengths were indicated in Science and, to a lesser extent, in Mathematics. Language skills were significant predictors of performance in all three core subjects. PIQ was the strongest predictor for Mathematics. For Science, both early language skills at 7 years and PIQ made significant contributions. Language impacts on the school performance of children with SLI, but differentially across subjects. English for these children is the most challenging of the core subjects, reflecting the high levels of language demand it incurs. Science is an area of relative strength and mathematics appears to be intermediate, arguably because some tasks in these subjects can be performed with less reliance on verbal processing. Many children with SLI do have the potential to reach or exceed educational targets that are set

  8. 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

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

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

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  12. A Critical Analysis of Anesthesiology Podcasts: Identifying Determinants of Success.

    Science.gov (United States)

    Singh, Devin; Alam, Fahad; Matava, Clyde

    2016-08-17

    Audio and video podcasts have gained popularity in recent years. Increasingly, podcasts are being used in the field of medicine as a tool to disseminate information. This format has multiple advantages including highly accessible creation tools, low distribution costs, and portability for the user. However, despite its ongoing use in medical education, there are no data describing factors associated with the success or quality of podcasts. The goal of the study was to assess the landscape of anesthesia podcasts in Canada and develop a methodology for evaluating the quality of the podcast. To achieve our objective, we identified the scope of podcasts in anesthesia specifically, constructed an algorithmic model for measuring success, and identified factors linked to both successful podcasts and a peer-review process. Independent reviewers performed a systematic search of anesthesia-related podcasts on iTunes Canada. Data and metrics recorded for each podcast included podcast's authorship, number posted, podcast series duration, target audience, topics, and social media presence. Descriptive statistics summarized mined data, and univariate analysis was used to identify factors associated with podcast success and a peer-review process. Twenty-two podcasts related to anesthesia were included in the final analysis. Less than a third (6/22=27%) were still active. The median longevity of the podcasts' series was just 13 months (interquartile range: 1-39 months). Anesthesiologists were the target audience for 77% of podcast series with clinical topics being most commonly addressed. We defined a novel algorithm for measuring success: Podcast Success Index. Factors associated with a high Podcast Success Index included podcasts targeting fellows (Spearman R=0.434; P=.04), inclusion of professional topics (Spearman R=0.456-0.603; P=.01-.03), and the use of Twitter as a means of social media (Spearman R=0.453;P=.03). In addition, more than two-thirds (16/22=73%) of podcasts

  13. 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)

  14. Identifying subgroups of patients using latent class analysis

    DEFF Research Database (Denmark)

    Nielsen, Anne Mølgaard; Kent, Peter; Hestbæk, Lise

    2017-01-01

    BACKGROUND: Heterogeneity in patients with low back pain (LBP) is well recognised and different approaches to subgrouping have been proposed. Latent Class Analysis (LCA) is a statistical technique that is increasingly being used to identify subgroups based on patient characteristics. However......, as LBP is a complex multi-domain condition, the optimal approach when using LCA is unknown. Therefore, this paper describes the exploration of two approaches to LCA that may help improve the identification of clinically relevant and interpretable LBP subgroups. METHODS: From 928 LBP patients consulting...... of statistical performance measures, qualitative evaluation of clinical interpretability (face validity) and a subgroup membership comparison. RESULTS: For the single-stage LCA, a model solution with seven patient subgroups was preferred, and for the two-stage LCA, a nine patient subgroup model. Both approaches...

  15. Cluster Analysis of Clinical Data Identifies Fibromyalgia Subgroups

    Science.gov (United States)

    Docampo, Elisa; Collado, Antonio; Escaramís, Geòrgia; Carbonell, Jordi; Rivera, Javier; Vidal, Javier; Alegre, José

    2013-01-01

    Introduction Fibromyalgia (FM) is mainly characterized by widespread pain and multiple accompanying symptoms, which hinder FM assessment and management. In order to reduce FM heterogeneity we classified clinical data into simplified dimensions that were used to define FM subgroups. Material and Methods 48 variables were evaluated in 1,446 Spanish FM cases fulfilling 1990 ACR FM criteria. A partitioning analysis was performed to find groups of variables similar to each other. Similarities between variables were identified and the variables were grouped into dimensions. This was performed in a subset of 559 patients, and cross-validated in the remaining 887 patients. For each sample and dimension, a composite index was obtained based on the weights of the variables included in the dimension. Finally, a clustering procedure was applied to the indexes, resulting in FM subgroups. Results Variables clustered into three independent dimensions: “symptomatology”, “comorbidities” and “clinical scales”. Only the two first dimensions were considered for the construction of FM subgroups. Resulting scores classified FM samples into three subgroups: low symptomatology and comorbidities (Cluster 1), high symptomatology and comorbidities (Cluster 2), and high symptomatology but low comorbidities (Cluster 3), showing differences in measures of disease severity. Conclusions We have identified three subgroups of FM samples in a large cohort of FM by clustering clinical data. Our analysis stresses the importance of family and personal history of FM comorbidities. Also, the resulting patient clusters could indicate different forms of the disease, relevant to future research, and might have an impact on clinical assessment. PMID:24098674

  16. Full core reactor analysis: Running Denovo on Jaguar

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  17. Whole-core analysis by 13C NMR

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. Identifying avian sources of faecal contamination using sterol analysis.

    Science.gov (United States)

    Devane, Megan L; Wood, David; Chappell, Andrew; Robson, Beth; Webster-Brown, Jenny; Gilpin, Brent J

    2015-10-01

    Discrimination of the source of faecal pollution in water bodies is an important step in the assessment and mitigation of public health risk. One tool for faecal source tracking is the analysis of faecal sterols which are present in faeces of animals in a range of distinctive ratios. Published ratios are able to discriminate between human and herbivore mammal faecal inputs but are of less value for identifying pollution from wildfowl, which can be a common cause of elevated bacterial indicators in rivers and streams. In this study, the sterol profiles of 50 avian-derived faecal specimens (seagulls, ducks and chickens) were examined alongside those of 57 ruminant faeces and previously published sterol profiles of human wastewater, chicken effluent and animal meatwork effluent. Two novel sterol ratios were identified as specific to avian faecal scats, which, when incorporated into a decision tree with human and herbivore mammal indicative ratios, were able to identify sterols from avian-polluted waterways. For samples where the sterol profile was not consistent with herbivore mammal or human pollution, avian pollution is indicated when the ratio of 24-ethylcholestanol/(24-ethylcholestanol + 24-ethylcoprostanol + 24-ethylepicoprostanol) is ≥0.4 (avian ratio 1) and the ratio of cholestanol/(cholestanol + coprostanol + epicoprostanol) is ≥0.5 (avian ratio 2). When avian pollution is indicated, further confirmation by targeted PCR specific markers can be employed if greater confidence in the pollution source is required. A 66% concordance between sterol ratios and current avian PCR markers was achieved when 56 water samples from polluted waterways were analysed.

  19. Social network analysis in identifying influential webloggers: A preliminary study

    Science.gov (United States)

    Hasmuni, Noraini; Sulaiman, Nor Intan Saniah; Zaibidi, Nerda Zura

    2014-12-01

    In recent years, second generation of internet-based services such as weblog has become an effective communication tool to publish information on the Web. Weblogs have unique characteristics that deserve users' attention. Some of webloggers have seen weblogs as appropriate medium to initiate and expand business. These webloggers or also known as direct profit-oriented webloggers (DPOWs) communicate and share knowledge with each other through social interaction. However, survivability is the main issue among DPOW. Frequent communication with influential webloggers is one of the way to keep survive as DPOW. This paper aims to understand the network structure and identify influential webloggers within the network. Proper understanding of the network structure can assist us in knowing how the information is exchanged among members and enhance survivability among DPOW. 30 DPOW were involved in this study. Degree centrality and betweenness centrality measurement in Social Network Analysis (SNA) were used to examine the strength relation and identify influential webloggers within the network. Thus, webloggers with the highest value of these measurements are considered as the most influential webloggers in the network.

  20. Social Network Analysis Identifies Key Participants in Conservation Development.

    Science.gov (United States)

    Farr, Cooper M; Reed, Sarah E; Pejchar, Liba

    2018-05-01

    Understanding patterns of participation in private lands conservation, which is often implemented voluntarily by individual citizens and private organizations, could improve its effectiveness at combating biodiversity loss. We used social network analysis (SNA) to examine participation in conservation development (CD), a private land conservation strategy that clusters houses in a small portion of a property while preserving the remaining land as protected open space. Using data from public records for six counties in Colorado, USA, we compared CD participation patterns among counties and identified actors that most often work with others to implement CDs. We found that social network characteristics differed among counties. The network density, or proportion of connections in the network, varied from fewer than 2 to nearly 15%, and was higher in counties with smaller populations and fewer CDs. Centralization, or the degree to which connections are held disproportionately by a few key actors, was not correlated strongly with any county characteristics. Network characteristics were not correlated with the prevalence of wildlife-friendly design features in CDs. The most highly connected actors were biological and geological consultants, surveyors, and engineers. Our work demonstrates a new application of SNA to land-use planning, in which CD network patterns are examined and key actors are identified. For better conservation outcomes of CD, we recommend using network patterns to guide strategies for outreach and information dissemination, and engaging with highly connected actor types to encourage widespread adoption of best practices for CD design and stewardship.

  1. Global Proteome Analysis Identifies Active Immunoproteasome Subunits in Human Platelets*

    Science.gov (United States)

    Klockenbusch, Cordula; Walsh, Geraldine M.; Brown, Lyda M.; Hoffman, Michael D.; Ignatchenko, Vladimir; Kislinger, Thomas; Kast, Juergen

    2014-01-01

    The discovery of new functions for platelets, particularly in inflammation and immunity, has expanded the role of these anucleate cell fragments beyond their primary hemostatic function. Here, four in-depth human platelet proteomic data sets were generated to explore potential new functions for platelets based on their protein content and this led to the identification of 2559 high confidence proteins. During a more detailed analysis, consistently high expression of the proteasome was discovered, and the composition and function of this complex, whose role in platelets has not been thoroughly investigated, was examined. Data set mining resulted in identification of nearly all members of the 26S proteasome in one or more data sets, except the β5 subunit. However, β5i, a component of the immunoproteasome, was identified. Biochemical analyses confirmed the presence of all catalytically active subunits of the standard 20S proteasome and immunoproteasome in human platelets, including β5, which was predominantly found in its precursor form. It was demonstrated that these components were assembled into the proteasome complex and that standard proteasome as well as immunoproteasome subunits were constitutively active in platelets. These findings suggest potential new roles for platelets in the immune system. For example, the immunoproteasome may be involved in major histocompatibility complex I (MHC I) peptide generation, as the MHC I machinery was also identified in our data sets. PMID:25146974

  2. Global proteome analysis identifies active immunoproteasome subunits in human platelets.

    Science.gov (United States)

    Klockenbusch, Cordula; Walsh, Geraldine M; Brown, Lyda M; Hoffman, Michael D; Ignatchenko, Vladimir; Kislinger, Thomas; Kast, Juergen

    2014-12-01

    The discovery of new functions for platelets, particularly in inflammation and immunity, has expanded the role of these anucleate cell fragments beyond their primary hemostatic function. Here, four in-depth human platelet proteomic data sets were generated to explore potential new functions for platelets based on their protein content and this led to the identification of 2559 high confidence proteins. During a more detailed analysis, consistently high expression of the proteasome was discovered, and the composition and function of this complex, whose role in platelets has not been thoroughly investigated, was examined. Data set mining resulted in identification of nearly all members of the 26S proteasome in one or more data sets, except the β5 subunit. However, β5i, a component of the immunoproteasome, was identified. Biochemical analyses confirmed the presence of all catalytically active subunits of the standard 20S proteasome and immunoproteasome in human platelets, including β5, which was predominantly found in its precursor form. It was demonstrated that these components were assembled into the proteasome complex and that standard proteasome as well as immunoproteasome subunits were constitutively active in platelets. These findings suggest potential new roles for platelets in the immune system. For example, the immunoproteasome may be involved in major histocompatibility complex I (MHC I) peptide generation, as the MHC I machinery was also identified in our data sets. © 2014 by The American Society for Biochemistry and Molecular Biology, Inc.

  3. 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

  4. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    Dubuisson, B.; Smolarz, A.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Nakao, Y.

    2002-01-01

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

  6. 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

  7. 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)

  8. 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

  9. 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

  10. Core and shielding analysis of the SCM-100

    International Nuclear Information System (INIS)

    Olson, A. P.

    2002-01-01

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

  11. 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...

  12. 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)

  13. 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...

  14. 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

  15. 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

  16. 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.)

  17. 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.)

  18. Geochemical analysis of core from a geothermal anomaly

    International Nuclear Information System (INIS)

    Haverslew, B.; Tammemagi, H.Y.

    1985-04-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. Identifying target processes for microbial electrosynthesis by elementary mode analysis.

    Science.gov (United States)

    Kracke, Frauke; Krömer, Jens O

    2014-12-30

    Microbial electrosynthesis and electro fermentation are techniques that aim to optimize microbial production of chemicals and fuels by regulating the cellular redox balance via interaction with electrodes. While the concept is known for decades major knowledge gaps remain, which make it hard to evaluate its biotechnological potential. Here we present an in silico approach to identify beneficial production processes for electro fermentation by elementary mode analysis. Since the fundamentals of electron transport between electrodes and microbes have not been fully uncovered yet, we propose different options and discuss their impact on biomass and product yields. For the first time 20 different valuable products were screened for their potential to show increased yields during anaerobic electrically enhanced fermentation. Surprisingly we found that an increase in product formation by electrical enhancement is not necessarily dependent on the degree of reduction of the product but rather the metabolic pathway it is derived from. We present a variety of beneficial processes with product yield increases of maximal 36% in reductive and 84% in oxidative fermentations and final theoretical product yields up to 100%. This includes compounds that are already produced at industrial scale such as succinic acid, lysine and diaminopentane as well as potential novel bio-commodities such as isoprene, para-hydroxybenzoic acid and para-aminobenzoic acid. Furthermore, it is shown that the way of electron transport has major impact on achievable biomass and product yields. The coupling of electron transport to energy conservation could be identified as crucial for most processes. This study introduces a powerful tool to determine beneficial substrate and product combinations for electro-fermentation. It also highlights that the maximal yield achievable by bio electrochemical techniques depends strongly on the actual electron transport mechanisms. Therefore it is of great importance to

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. 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

  4. 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

  5. Behavioral metabolomics analysis identifies novel neurochemical signatures in methamphetamine sensitization

    Science.gov (United States)

    Adkins, Daniel E.; McClay, Joseph L.; Vunck, Sarah A.; Batman, Angela M.; Vann, Robert E.; Clark, Shaunna L.; Souza, Renan P.; Crowley, James J.; Sullivan, Patrick F.; van den Oord, Edwin J.C.G.; Beardsley, Patrick M.

    2014-01-01

    Behavioral sensitization has been widely studied in animal models and is theorized to reflect neural modifications associated with human psychostimulant addiction. While the mesolimbic dopaminergic pathway is known to play a role, the neurochemical mechanisms underlying behavioral sensitization remain incompletely understood. In the present study, we conducted the first metabolomics analysis to globally characterize neurochemical differences associated with behavioral sensitization. Methamphetamine-induced sensitization measures were generated by statistically modeling longitudinal activity data for eight inbred strains of mice. Subsequent to behavioral testing, nontargeted liquid and gas chromatography-mass spectrometry profiling was performed on 48 brain samples, yielding 301 metabolite levels per sample after quality control. Association testing between metabolite levels and three primary dimensions of behavioral sensitization (total distance, stereotypy and margin time) showed four robust, significant associations at a stringent metabolome-wide significance threshold (false discovery rate < 0.05). Results implicated homocarnosine, a dipeptide of GABA and histidine, in total distance sensitization, GABA metabolite 4-guanidinobutanoate and pantothenate in stereotypy sensitization, and myo-inositol in margin time sensitization. Secondary analyses indicated that these associations were independent of concurrent methamphetamine levels and, with the exception of the myo-inositol association, suggest a mechanism whereby strain-based genetic variation produces specific baseline neurochemical differences that substantially influence the magnitude of MA-induced sensitization. These findings demonstrate the utility of mouse metabolomics for identifying novel biomarkers, and developing more comprehensive neurochemical models, of psychostimulant sensitization. PMID:24034544

  6. A Sensitivity Analysis Approach to Identify Key Environmental Performance Factors

    Directory of Open Access Journals (Sweden)

    Xi Yu

    2014-01-01

    Full Text Available Life cycle assessment (LCA is widely used in design phase to reduce the product’s environmental impacts through the whole product life cycle (PLC during the last two decades. The traditional LCA is restricted to assessing the environmental impacts of a product and the results cannot reflect the effects of changes within the life cycle. In order to improve the quality of ecodesign, it is a growing need to develop an approach which can reflect the changes between the design parameters and product’s environmental impacts. A sensitivity analysis approach based on LCA and ecodesign is proposed in this paper. The key environmental performance factors which have significant influence on the products’ environmental impacts can be identified by analyzing the relationship between environmental impacts and the design parameters. Users without much environmental knowledge can use this approach to determine which design parameter should be first considered when (redesigning a product. A printed circuit board (PCB case study is conducted; eight design parameters are chosen to be analyzed by our approach. The result shows that the carbon dioxide emission during the PCB manufacture is highly sensitive to the area of PCB panel.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Tukiran-Surbakti

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

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

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan

    1989-01-01

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

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

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

    International Nuclear Information System (INIS)

    Grzesik, B; Stepien, M; Jez, R

    2010-01-01

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

  20. 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…

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

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Dumitrache, I.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-04-01

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

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

    International Nuclear Information System (INIS)

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

    1987-05-01

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

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

    International Nuclear Information System (INIS)

    Khan, R.

    2010-01-01

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

  5. 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)

  6. 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)

  7. 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)

  8. 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)

  9. 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.

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

    International Nuclear Information System (INIS)

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

    1993-07-01

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

  11. 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

  12. Identifying changes in the support networks of end-of-life carers using social network analysis.

    Science.gov (United States)

    Leonard, Rosemary; Horsfall, Debbie; Noonan, Kerrie

    2015-06-01

    End-of-life caring is often associated with reduced social networks for both the dying person and for the carer. However, those adopting a community participation and development approach, see the potential for the expansion and strengthening of networks. This paper uses Knox, Savage and Harvey's definitions of three generations social network analysis to analyse the caring networks of people with a terminal illness who are being cared for at home and identifies changes in these caring networks that occurred over the period of caring. Participatory network mapping of initial and current networks was used in nine focus groups. The analysis used key concepts from social network analysis (size, density, transitivity, betweenness and local clustering) together with qualitative analyses of the group's reflections on the maps. The results showed an increase in the size of the networks and that ties between the original members of the network strengthened. The qualitative data revealed the importance between core and peripheral network members and the diverse contributions of the network members. The research supports the value of third generation social network analysis and the potential for end-of-life caring to build social capital. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  13. 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.

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

    International Nuclear Information System (INIS)

    Ravnik, M.

    1988-11-01

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

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

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

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

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

  17. 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.

  18. Staff Performance Analysis: A Method for Identifying Brigade Staff Tasks

    National Research Council Canada - National Science Library

    Ford, Laura

    1997-01-01

    ... members of conventional mounted brigade staff. Initial analysis of performance requirements in existing documentation revealed that the performance specifications were not sufficiently detailed for brigade battle staffs...

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

    Science.gov (United States)

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

    2018-02-07

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

  20. Structural identifiability analysis of a cardiovascular system model.

    Science.gov (United States)

    Pironet, Antoine; Dauby, Pierre C; Chase, J Geoffrey; Docherty, Paul D; Revie, James A; Desaive, Thomas

    2016-05-01

    The six-chamber cardiovascular system model of Burkhoff and Tyberg has been used in several theoretical and experimental studies. However, this cardiovascular system model (and others derived from it) are not identifiable from any output set. In this work, two such cases of structural non-identifiability are first presented. These cases occur when the model output set only contains a single type of information (pressure or volume). A specific output set is thus chosen, mixing pressure and volume information and containing only a limited number of clinically available measurements. Then, by manipulating the model equations involving these outputs, it is demonstrated that the six-chamber cardiovascular system model is structurally globally identifiable. A further simplification is made, assuming known cardiac valve resistances. Because of the poor practical identifiability of these four parameters, this assumption is usual. Under this hypothesis, the six-chamber cardiovascular system model is structurally identifiable from an even smaller dataset. As a consequence, parameter values computed from limited but well-chosen datasets are theoretically unique. This means that the parameter identification procedure can safely be performed on the model from such a well-chosen dataset. Thus, the model may be considered suitable for use in diagnosis. Copyright © 2016 IPEM. Published by Elsevier Ltd. All rights reserved.

  1. 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

  2. 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...

  3. Metatranscriptomic analysis of diverse microbial communities reveals core metabolic pathways and microbiome-specific functionality.

    Science.gov (United States)

    Jiang, Yue; Xiong, Xuejian; Danska, Jayne; Parkinson, John

    2016-01-12

    Metatranscriptomics is emerging as a powerful technology for the functional characterization of complex microbial communities (microbiomes). Use of unbiased RNA-sequencing can reveal both the taxonomic composition and active biochemical functions of a complex microbial community. However, the lack of established reference genomes, computational tools and pipelines make analysis and interpretation of these datasets challenging. Systematic studies that compare data across microbiomes are needed to demonstrate the ability of such pipelines to deliver biologically meaningful insights on microbiome function. Here, we apply a standardized analytical pipeline to perform a comparative analysis of metatranscriptomic data from diverse microbial communities derived from mouse large intestine, cow rumen, kimchi culture, deep-sea thermal vent and permafrost. Sequence similarity searches allowed annotation of 19 to 76% of putative messenger RNA (mRNA) reads, with the highest frequency in the kimchi dataset due to its relatively low complexity and availability of closely related reference genomes. Metatranscriptomic datasets exhibited distinct taxonomic and functional signatures. From a metabolic perspective, we identified a common core of enzymes involved in amino acid, energy and nucleotide metabolism and also identified microbiome-specific pathways such as phosphonate metabolism (deep sea) and glycan degradation pathways (cow rumen). Integrating taxonomic and functional annotations within a novel visualization framework revealed the contribution of different taxa to metabolic pathways, allowing the identification of taxa that contribute unique functions. The application of a single, standard pipeline confirms that the rich taxonomic and functional diversity observed across microbiomes is not simply an artefact of different analysis pipelines but instead reflects distinct environmental influences. At the same time, our findings show how microbiome complexity and availability of

  4. 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.)

  5. 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.

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

    International Nuclear Information System (INIS)

    Whetton, P.A.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Zhu Huanjun; Xu Yijun

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Wieselquist, William; Ferroukhi, Hakim; Bernatowicz, Kinga

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    1985-05-01

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

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

    International Nuclear Information System (INIS)

    Endiah-Puji-Hastuti

    2003-01-01

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

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

    DEFF Research Database (Denmark)

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

    2016-01-01

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

  14. 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.

  15. 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

  16. 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

  17. Building waste management core indicators through Spatial Material Flow Analysis: Net recovery and transport intensity indexes

    Energy Technology Data Exchange (ETDEWEB)

    Font Vivanco, David, E-mail: font@cml.leidenuniv.nl [Institut de Ciencia i Tecnologia Ambientals (ICTA), Departament d' Enginyeria Quimica, Universitat Autonoma de Barcelona (UAB), 08193 Bellaterra, Barcelona (Spain); Institute of Environmental Sciences (CML), Leiden University, P.O. Box 9518, 2300 RA Leiden (Netherlands); Puig Ventosa, Ignasi [ENT Environment and Management, Carrer Sant Joan 39, First Floor, 08800 Vilanova i la Geltru, Barcelona (Spain); Gabarrell Durany, Xavier [Institut de Ciencia i Tecnologia Ambientals (ICTA), Departament d' Enginyeria Quimica, Universitat Autonoma de Barcelona (UAB), 08193 Bellaterra, Barcelona (Spain)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Sustainability and proximity principles have a key role in waste management. Black-Right-Pointing-Pointer Core indicators are needed in order to quantify and evaluate them. Black-Right-Pointing-Pointer A systematic, step-by-step approach is developed in this study for their development. Black-Right-Pointing-Pointer Transport may play a significant role in terms of environmental and economic costs. Black-Right-Pointing-Pointer Policy action is required in order to advance in the consecution of these principles. - Abstract: In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy

  18. Building waste management core indicators through Spatial Material Flow Analysis: Net recovery and transport intensity indexes

    International Nuclear Information System (INIS)

    Font Vivanco, David; Puig Ventosa, Ignasi; Gabarrell Durany, Xavier

    2012-01-01

    Highlights: ► Sustainability and proximity principles have a key role in waste management. ► Core indicators are needed in order to quantify and evaluate them. ► A systematic, step-by-step approach is developed in this study for their development. ► Transport may play a significant role in terms of environmental and economic costs. ► Policy action is required in order to advance in the consecution of these principles. - Abstract: In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy prioritization. Moreover, this methodological approach permits scenario building, which could be useful in assessing the outcomes of

  19. Gene expression analysis identifies global gene dosage sensitivity in cancer

    DEFF Research Database (Denmark)

    Fehrmann, Rudolf S. N.; Karjalainen, Juha M.; Krajewska, Malgorzata

    2015-01-01

    Many cancer-associated somatic copy number alterations (SCNAs) are known. Currently, one of the challenges is to identify the molecular downstream effects of these variants. Although several SCNAs are known to change gene expression levels, it is not clear whether each individual SCNA affects gen...

  20. Potential bacterial core species associated with digital dermatitis in cattle herds identified by molecular profiling of interdigital skin samples

    DEFF Research Database (Denmark)

    Weiss Nielsen, Martin; Strube, Mikael Lenz; Isbrand, Anastasia

    2016-01-01

    of different molecular methods. Deep sequencing of the 16S rRNA gene variable regions V1–V2 showed that Treponema, Mycoplasma, Fusobacterium and Porphyromonas were the genera best differentiating the DD samples from the controls. Additional deep sequencing analysis of the most abundant genus, Treponema...... in the epidermal lesions and were present in only a subset of samples. RT-qPCR analysis showed that treponemes were also actively expressing a panel of virulence factors at the site of infection. Our results further support the hypothesis that species belonging to the genus Treponema are major pathogens of DD...... and also provide sufficient clues to motivate additional research into the role of M. fermentans, F. necrophorum and P. levii in the etiology of DD....

  1. 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.)

  2. Genomic analysis identifies masqueraders of full-term cerebral palsy.

    Science.gov (United States)

    Takezawa, Yusuke; Kikuchi, Atsuo; Haginoya, Kazuhiro; Niihori, Tetsuya; Numata-Uematsu, Yurika; Inui, Takehiko; Yamamura-Suzuki, Saeko; Miyabayashi, Takuya; Anzai, Mai; Suzuki-Muromoto, Sato; Okubo, Yukimune; Endo, Wakaba; Togashi, Noriko; Kobayashi, Yasuko; Onuma, Akira; Funayama, Ryo; Shirota, Matsuyuki; Nakayama, Keiko; Aoki, Yoko; Kure, Shigeo

    2018-05-01

    Cerebral palsy is a common, heterogeneous neurodevelopmental disorder that causes movement and postural disabilities. Recent studies have suggested genetic diseases can be misdiagnosed as cerebral palsy. We hypothesized that two simple criteria, that is, full-term births and nonspecific brain MRI findings, are keys to extracting masqueraders among cerebral palsy cases due to the following: (1) preterm infants are susceptible to multiple environmental factors and therefore demonstrate an increased risk of cerebral palsy and (2) brain MRI assessment is essential for excluding environmental causes and other particular disorders. A total of 107 patients-all full-term births-without specific findings on brain MRI were identified among 897 patients diagnosed with cerebral palsy who were followed at our center. DNA samples were available for 17 of the 107 cases for trio whole-exome sequencing and array comparative genomic hybridization. We prioritized variants in genes known to be relevant in neurodevelopmental diseases and evaluated their pathogenicity according to the American College of Medical Genetics guidelines. Pathogenic/likely pathogenic candidate variants were identified in 9 of 17 cases (52.9%) within eight genes: CTNNB1 , CYP2U1 , SPAST , GNAO1 , CACNA1A , AMPD2 , STXBP1 , and SCN2A . Five identified variants had previously been reported. No pathogenic copy number variations were identified. The AMPD2 missense variant and the splice-site variants in CTNNB1 and AMPD2 were validated by in vitro functional experiments. The high rate of detecting causative genetic variants (52.9%) suggests that patients diagnosed with cerebral palsy in full-term births without specific MRI findings may include genetic diseases masquerading as cerebral palsy.

  3. Association analysis identifies ZNF750 regulatory variants in psoriasis

    Directory of Open Access Journals (Sweden)

    Birnbaum Ramon Y

    2011-12-01

    Full Text Available Abstract Background Mutations in the ZNF750 promoter and coding regions have been previously associated with Mendelian forms of psoriasis and psoriasiform dermatitis. ZNF750 encodes a putative zinc finger transcription factor that is highly expressed in keratinocytes and represents a candidate psoriasis gene. Methods We examined whether ZNF750 variants were associated with psoriasis in a large case-control population. We sequenced the promoter and exon regions of ZNF750 in 716 Caucasian psoriasis cases and 397 Caucasian controls. Results We identified a total of 47 variants, including 38 rare variants of which 35 were novel. Association testing identified two ZNF750 haplotypes associated with psoriasis (p ZNF750 promoter and 5' UTR variants displayed a 35-55% reduction of ZNF750 promoter activity, consistent with the promoter activity reduction seen in a Mendelian psoriasis family with a ZNF750 promoter variant. However, the rare promoter and 5' UTR variants identified in this study did not strictly segregate with the psoriasis phenotype within families. Conclusions Two haplotypes of ZNF750 and rare 5' regulatory variants of ZNF750 were found to be associated with psoriasis. These rare 5' regulatory variants, though not causal, might serve as a genetic modifier of psoriasis.

  4. 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.

  5. 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

  6. 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

  7. Using Factor Analysis to Identify Topic Preferences Within MBA Courses

    Directory of Open Access Journals (Sweden)

    Earl Chrysler

    2003-02-01

    Full Text Available This study demonstrates the role of a principal components factor analysis in conducting a gap analysis as to the desired characteristics of business alumni. Typically, gap analyses merely compare the emphases that should be given to areas of inquiry with perceptions of actual emphases. As a result, the focus is upon depth of coverage. A neglected area in need of investigation is the breadth of topic dimensions and their differences between the normative (should offer and the descriptive (actually offer. The implications of factor structures, as well as traditional gap analyses, are developed and discussed in the context of outcomes assessment.

  8. 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)

  9. Proteomic analysis of cell lines to identify the irinotecan resistance ...

    Indian Academy of Sciences (India)

    MADHU

    was selected from the wild-type LoVo cell line by chronic exposure to irinotecan ... dose–effect curves of anticancer drugs were drawn on semilogarithm .... alcohol metabolites daunorubicinol (Forrest and Gonzalez. 2000; Mordente et al. ..... Chen L, Huang C and Wei Y 2007 Proteomic analysis of liver cancer cells treated ...

  10. DRIS Analysis Identifies a Common Potassium Imbalance in Sweetgum Plantations

    Science.gov (United States)

    Mark D. Coleman; S.X. Chang; D.J. Robison

    2003-01-01

    DRIS (Diagnosis and Recommendation Integrated System) analysis was applied to fast-growing sweetgum (Liquidambar styraciflua L.) plantations in the southeast United States as a tool for nutrient diagnosis and fertilizer recommendations. First, standard foliar nutrient ratios for nitrogen (N), phosphorus (P), potassium (K), calcium (Ca), and...

  11. 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.)

  12. 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.

  13. 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

  14. 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)

  15. 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)

  16. Association analysis identifies 65 new breast cancer risk loci

    OpenAIRE

    Michailidou, Kyriaki; Lindström, Sara; Dennis, Joe; Beesley, Jonathan; Hui, Shirley; Kar, Siddhartha; Lemaçon, Audrey; Soucy, Penny; Glubb, Dylan; Rostamianfar, Asha; Bolla, Manjeet K; Wang, Qin; Tyrer, Jonathan; Dicks, Ed; Lee, Andrew

    2017-01-01

    Breast cancer risk is influenced by rare coding variants in susceptibility genes, such as BRCA1, and many common, mostly non-coding variants. However, much of the genetic contribution to breast cancer risk remains unknown. Here we report the results of a genome-wide association study of breast cancer in 122,977 cases and 105,974 controls of European ancestry and 14,068 cases and 13,104 controls of East Asian ancestry. We identified 65 new loci that are associated with overall breast cancer ri...

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. Structural and practical identifiability analysis of S-system.

    Science.gov (United States)

    Zhan, Choujun; Li, Benjamin Yee Shing; Yeung, Lam Fat

    2015-12-01

    In the field of systems biology, biological reaction networks are usually modelled by ordinary differential equations. A sub-class, the S-systems representation, is a widely used form of modelling. Existing S-systems identification techniques assume that the system itself is always structurally identifiable. However, due to practical limitations, biological reaction networks are often only partially measured. In addition, the captured data only covers a limited trajectory, therefore data can only be considered as a local snapshot of the system responses with respect to the complete set of state trajectories over the entire state space. Hence the estimated model can only reflect partial system dynamics and may not be unique. To improve the identification quality, the structural and practical identifiablility of S-system are studied. The S-system is shown to be identifiable under a set of assumptions. Then, an application on yeast fermentation pathway was conducted. Two case studies were chosen; where the first case is based on a larger state trajectories and the second case is based on a smaller one. By expanding the dataset which span a relatively larger state space, the uncertainty of the estimated system can be reduced. The results indicated that initial concentration is related to the practical identifiablity.

  6. 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)

  7. Association analysis identifies 65 new breast cancer risk loci

    Science.gov (United States)

    Lemaçon, Audrey; Soucy, Penny; Glubb, Dylan; Rostamianfar, Asha; Bolla, Manjeet K.; Wang, Qin; Tyrer, Jonathan; Dicks, Ed; Lee, Andrew; Wang, Zhaoming; Allen, Jamie; Keeman, Renske; Eilber, Ursula; French, Juliet D.; Chen, Xiao Qing; Fachal, Laura; McCue, Karen; McCart Reed, Amy E.; Ghoussaini, Maya; Carroll, Jason; Jiang, Xia; Finucane, Hilary; Adams, Marcia; Adank, Muriel A.; Ahsan, Habibul; Aittomäki, Kristiina; Anton-Culver, Hoda; Antonenkova, Natalia N.; Arndt, Volker; Aronson, Kristan J.; Arun, Banu; Auer, Paul L.; Bacot, François; Barrdahl, Myrto; Baynes, Caroline; Beckmann, Matthias W.; Behrens, Sabine; Benitez, Javier; Bermisheva, Marina; Bernstein, Leslie; Blomqvist, Carl; Bogdanova, Natalia V.; Bojesen, Stig E.; Bonanni, Bernardo; Børresen-Dale, Anne-Lise; Brand, Judith S.; Brauch, Hiltrud; Brennan, Paul; Brenner, Hermann; Brinton, Louise; Broberg, Per; Brock, Ian W.; Broeks, Annegien; Brooks-Wilson, Angela; Brucker, Sara Y.; Brüning, Thomas; Burwinkel, Barbara; Butterbach, Katja; Cai, Qiuyin; Cai, Hui; Caldés, Trinidad; Canzian, Federico; Carracedo, Angel; Carter, Brian D.; Castelao, Jose E.; Chan, Tsun L.; Cheng, Ting-Yuan David; Chia, Kee Seng; Choi, Ji-Yeob; Christiansen, Hans; Clarke, Christine L.; Collée, Margriet; Conroy, Don M.; Cordina-Duverger, Emilie; Cornelissen, Sten; Cox, David G; Cox, Angela; Cross, Simon S.; Cunningham, Julie M.; Czene, Kamila; Daly, Mary B.; Devilee, Peter; Doheny, Kimberly F.; Dörk, Thilo; dos-Santos-Silva, Isabel; Dumont, Martine; Durcan, Lorraine; Dwek, Miriam; Eccles, Diana M.; Ekici, Arif B.; Eliassen, A. Heather; Ellberg, Carolina; Elvira, Mingajeva; Engel, Christoph; Eriksson, Mikael; Fasching, Peter A.; Figueroa, Jonine; Flesch-Janys, Dieter; Fletcher, Olivia; Flyger, Henrik; Fritschi, Lin; Gaborieau, Valerie; Gabrielson, Marike; Gago-Dominguez, Manuela; Gao, Yu-Tang; Gapstur, Susan M.; García-Sáenz, José A.; Gaudet, Mia M.; Georgoulias, Vassilios; Giles, Graham G.; Glendon, Gord; Goldberg, Mark S.; Goldgar, David E.; González-Neira, Anna; Grenaker Alnæs, Grethe I.; Grip, Mervi; Gronwald, Jacek; Grundy, Anne; Guénel, Pascal; Haeberle, Lothar; Hahnen, Eric; Haiman, Christopher A.; Håkansson, Niclas; Hamann, Ute; Hamel, Nathalie; Hankinson, Susan; Harrington, Patricia; Hart, Steven N.; Hartikainen, Jaana M.; Hartman, Mikael; Hein, Alexander; Heyworth, Jane; Hicks, Belynda; Hillemanns, Peter; Ho, Dona N.; Hollestelle, Antoinette; Hooning, Maartje J.; Hoover, Robert N.; Hopper, John L.; Hou, Ming-Feng; Hsiung, Chia-Ni; Huang, Guanmengqian; Humphreys, Keith; Ishiguro, Junko; Ito, Hidemi; Iwasaki, Motoki; Iwata, Hiroji; Jakubowska, Anna; Janni, Wolfgang; John, Esther M.; Johnson, Nichola; Jones, Kristine; Jones, Michael; Jukkola-Vuorinen, Arja; Kaaks, Rudolf; Kabisch, Maria; Kaczmarek, Katarzyna; Kang, Daehee; Kasuga, Yoshio; Kerin, Michael J.; Khan, Sofia; Khusnutdinova, Elza; Kiiski, Johanna I.; Kim, Sung-Won; Knight, Julia A.; Kosma, Veli-Matti; Kristensen, Vessela N.; Krüger, Ute; Kwong, Ava; Lambrechts, Diether; Marchand, Loic Le; Lee, Eunjung; Lee, Min Hyuk; Lee, Jong Won; Lee, Chuen Neng; Lejbkowicz, Flavio; Li, Jingmei; Lilyquist, Jenna; Lindblom, Annika; Lissowska, Jolanta; Lo, Wing-Yee; Loibl, Sibylle; Long, Jirong; Lophatananon, Artitaya; Lubinski, Jan; Luccarini, Craig; Lux, Michael P.; Ma, Edmond S.K.; MacInnis, Robert J.; Maishman, Tom; Makalic, Enes; Malone, Kathleen E; Kostovska, Ivana Maleva; Mannermaa, Arto; Manoukian, Siranoush; Manson, JoAnn E.; Margolin, Sara; Mariapun, Shivaani; Martinez, Maria Elena; Matsuo, Keitaro; Mavroudis, Dimitrios; McKay, James; McLean, Catriona; Meijers-Heijboer, Hanne; Meindl, Alfons; Menéndez, Primitiva; Menon, Usha; Meyer, Jeffery; Miao, Hui; Miller, Nicola; Mohd Taib, Nur Aishah; Muir, Kenneth; Mulligan, Anna Marie; Mulot, Claire; Neuhausen, Susan L.; Nevanlinna, Heli; Neven, Patrick; Nielsen, Sune F.; Noh, Dong-Young; Nordestgaard, Børge G.; Norman, Aaron; Olopade, Olufunmilayo I.; Olson, Janet E.; Olsson, Håkan; Olswold, Curtis; Orr, Nick; Pankratz, V. Shane; Park, Sue K.; Park-Simon, Tjoung-Won; Lloyd, Rachel; Perez, Jose I.A.; Peterlongo, Paolo; Peto, Julian; Phillips, Kelly-Anne; Pinchev, Mila; Plaseska-Karanfilska, Dijana; Prentice, Ross; Presneau, Nadege; Prokofieva, Darya; Pugh, Elizabeth; Pylkäs, Katri; Rack, Brigitte; Radice, Paolo; Rahman, Nazneen; Rennert, Gadi; Rennert, Hedy S.; Rhenius, Valerie; Romero, Atocha; Romm, Jane; Ruddy, Kathryn J; Rüdiger, Thomas; Rudolph, Anja; Ruebner, Matthias; Rutgers, Emiel J. Th.; Saloustros, Emmanouil; Sandler, Dale P.; Sangrajrang, Suleeporn; Sawyer, Elinor J.; Schmidt, Daniel F.; Schmutzler, Rita K.; Schneeweiss, Andreas; Schoemaker, Minouk J.; Schumacher, Fredrick; Schürmann, Peter; Scott, Rodney J.; Scott, Christopher; Seal, Sheila; Seynaeve, Caroline; Shah, Mitul; Sharma, Priyanka; Shen, Chen-Yang; Sheng, Grace; Sherman, Mark E.; Shrubsole, Martha J.; Shu, Xiao-Ou; Smeets, Ann; Sohn, Christof; Southey, Melissa C.; Spinelli, John J.; Stegmaier, Christa; Stewart-Brown, Sarah; Stone, Jennifer; Stram, Daniel O.; Surowy, Harald; Swerdlow, Anthony; Tamimi, Rulla; Taylor, Jack A.; Tengström, Maria; Teo, Soo H.; Terry, Mary Beth; Tessier, Daniel C.; Thanasitthichai, Somchai; Thöne, Kathrin; Tollenaar, Rob A.E.M.; Tomlinson, Ian; Tong, Ling; Torres, Diana; Truong, Thérèse; Tseng, Chiu-chen; Tsugane, Shoichiro; Ulmer, Hans-Ulrich; Ursin, Giske; Untch, Michael; Vachon, Celine; van Asperen, Christi J.; Van Den Berg, David; van den Ouweland, Ans M.W.; van der Kolk, Lizet; van der Luijt, Rob B.; Vincent, Daniel; Vollenweider, Jason; Waisfisz, Quinten; Wang-Gohrke, Shan; Weinberg, Clarice R.; Wendt, Camilla; Whittemore, Alice S.; Wildiers, Hans; Willett, Walter; Winqvist, Robert; Wolk, Alicja; Wu, Anna H.; Xia, Lucy; Yamaji, Taiki; Yang, Xiaohong R.; Yip, Cheng Har; Yoo, Keun-Young; Yu, Jyh-Cherng; Zheng, Wei; Zheng, Ying; Zhu, Bin; Ziogas, Argyrios; Ziv, Elad; Lakhani, Sunil R.; Antoniou, Antonis C.; Droit, Arnaud; Andrulis, Irene L.; Amos, Christopher I.; Couch, Fergus J.; Pharoah, Paul D.P.; Chang-Claude, Jenny; Hall, Per; Hunter, David J.; Milne, Roger L.; García-Closas, Montserrat; Schmidt, Marjanka K.; Chanock, Stephen J.; Dunning, Alison M.; Edwards, Stacey L.; Bader, Gary D.; Chenevix-Trench, Georgia; Simard, Jacques; Kraft, Peter; Easton, Douglas F.

    2017-01-01

    Breast cancer risk is influenced by rare coding variants in susceptibility genes such as BRCA1 and many common, mainly non-coding variants. However, much of the genetic contribution to breast cancer risk remains unknown. We report results from a genome-wide association study (GWAS) of breast cancer in 122,977 cases and 105,974 controls of European ancestry and 14,068 cases and 13,104 controls of East Asian ancestry1. We identified 65 new loci associated with overall breast cancer at pcancer due to all SNPs in regulatory features was 2-5-fold enriched relative to the genome-wide average, with strong enrichment for particular transcription factor binding sites. These results provide further insight into genetic susceptibility to breast cancer and will improve the utility of genetic risk scores for individualized screening and prevention. PMID:29059683

  8. Association analysis identifies 65 new breast cancer risk loci

    DEFF Research Database (Denmark)

    Michailidou, Kyriaki; Lindström, Sara; Dennis, Joe

    2017-01-01

    Breast cancer risk is influenced by rare coding variants in susceptibility genes, such as BRCA1, and many common, mostly non-coding variants. However, much of the genetic contribution to breast cancer risk remains unknown. Here we report the results of a genome-wide association study of breast...... cancer in 122,977 cases and 105,974 controls of European ancestry and 14,068 cases and 13,104 controls of East Asian ancestry. We identified 65 new loci that are associated with overall breast cancer risk at P risk single-nucleotide polymorphisms in these loci fall......-nucleotide polymorphisms in regulatory features was 2-5-fold enriched relative to the genome-wide average, with strong enrichment for particular transcription factor binding sites. These results provide further insight into genetic susceptibility to breast cancer and will improve the use of genetic risk scores...

  9. 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)

  10. 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

  11. Building waste management core indicators through Spatial Material Flow Analysis: net recovery and transport intensity indexes.

    Science.gov (United States)

    Font Vivanco, David; Puig Ventosa, Ignasi; Gabarrell Durany, Xavier

    2012-12-01

    In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy prioritization. Moreover, this methodological approach permits scenario building, which could be useful in assessing the outcomes of hypothetical scenarios, thus proving its adequacy for strategic planning. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-06

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

  13. 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

  14. 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....

  15. 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…

  16. 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

  17. 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

  18. Methodology for reactor core physics analysis - part 2; Metodologia de analise fisica do nucleo - etapa 2

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-01

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

  19. 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.

  20. Probabilistic analysis for identifying the driving force of protein folding

    Science.gov (United States)

    Tokunaga, Yoshihiko; Yamamori, Yu; Matubayasi, Nobuyuki

    2018-03-01

    Toward identifying the driving force of protein folding, energetics was analyzed in water for Trp-cage (20 residues), protein G (56 residues), and ubiquitin (76 residues) at their native (folded) and heat-denatured (unfolded) states. All-atom molecular dynamics simulation was conducted, and the hydration effect was quantified by the solvation free energy. The free-energy calculation was done by employing the solution theory in the energy representation, and it was seen that the sum of the protein intramolecular (structural) energy and the solvation free energy is more favorable for a folded structure than for an unfolded one generated by heat. Probabilistic arguments were then developed to determine which of the electrostatic, van der Waals, and excluded-volume components of the interactions in the protein-water system governs the relative stabilities between the folded and unfolded structures. It was found that the electrostatic interaction does not correspond to the preference order of the two structures. The van der Waals and excluded-volume components were shown, on the other hand, to provide the right order of preference at probabilities of almost unity, and it is argued that a useful modeling of protein folding is possible on the basis of the excluded-volume effect.

  1. Association analysis identifies 65 new breast cancer risk loci.

    Science.gov (United States)

    Michailidou, Kyriaki; Lindström, Sara; Dennis, Joe; Beesley, Jonathan; Hui, Shirley; Kar, Siddhartha; Lemaçon, Audrey; Soucy, Penny; Glubb, Dylan; Rostamianfar, Asha; Bolla, Manjeet K; Wang, Qin; Tyrer, Jonathan; Dicks, Ed; Lee, Andrew; Wang, Zhaoming; Allen, Jamie; Keeman, Renske; Eilber, Ursula; French, Juliet D; Qing Chen, Xiao; Fachal, Laura; McCue, Karen; McCart Reed, Amy E; Ghoussaini, Maya; Carroll, Jason S; Jiang, Xia; Finucane, Hilary; Adams, Marcia; Adank, Muriel A; Ahsan, Habibul; Aittomäki, Kristiina; Anton-Culver, Hoda; Antonenkova, Natalia N; Arndt, Volker; Aronson, Kristan J; Arun, Banu; Auer, Paul L; Bacot, François; Barrdahl, Myrto; Baynes, Caroline; Beckmann, Matthias W; Behrens, Sabine; Benitez, Javier; Bermisheva, Marina; Bernstein, Leslie; Blomqvist, Carl; Bogdanova, Natalia V; Bojesen, Stig E; Bonanni, Bernardo; Børresen-Dale, Anne-Lise; Brand, Judith S; Brauch, Hiltrud; Brennan, Paul; Brenner, Hermann; Brinton, Louise; Broberg, Per; Brock, Ian W; Broeks, Annegien; Brooks-Wilson, Angela; Brucker, Sara Y; Brüning, Thomas; Burwinkel, Barbara; Butterbach, Katja; Cai, Qiuyin; Cai, Hui; Caldés, Trinidad; Canzian, Federico; Carracedo, Angel; Carter, Brian D; Castelao, Jose E; Chan, Tsun L; David Cheng, Ting-Yuan; Seng Chia, Kee; Choi, Ji-Yeob; Christiansen, Hans; Clarke, Christine L; Collée, Margriet; Conroy, Don M; Cordina-Duverger, Emilie; Cornelissen, Sten; Cox, David G; Cox, Angela; Cross, Simon S; Cunningham, Julie M; Czene, Kamila; Daly, Mary B; Devilee, Peter; Doheny, Kimberly F; Dörk, Thilo; Dos-Santos-Silva, Isabel; Dumont, Martine; Durcan, Lorraine; Dwek, Miriam; Eccles, Diana M; Ekici, Arif B; Eliassen, A Heather; Ellberg, Carolina; Elvira, Mingajeva; Engel, Christoph; Eriksson, Mikael; Fasching, Peter A; Figueroa, Jonine; Flesch-Janys, Dieter; Fletcher, Olivia; Flyger, Henrik; Fritschi, Lin; Gaborieau, Valerie; Gabrielson, Marike; Gago-Dominguez, Manuela; Gao, Yu-Tang; Gapstur, Susan M; García-Sáenz, José A; Gaudet, Mia M; Georgoulias, Vassilios; Giles, Graham G; Glendon, Gord; Goldberg, Mark S; Goldgar, David E; González-Neira, Anna; Grenaker Alnæs, Grethe I; Grip, Mervi; Gronwald, Jacek; Grundy, Anne; Guénel, Pascal; Haeberle, Lothar; Hahnen, Eric; Haiman, Christopher A; Håkansson, Niclas; Hamann, Ute; Hamel, Nathalie; Hankinson, Susan; Harrington, Patricia; Hart, Steven N; Hartikainen, Jaana M; Hartman, Mikael; Hein, Alexander; Heyworth, Jane; Hicks, Belynda; Hillemanns, Peter; Ho, Dona N; Hollestelle, Antoinette; Hooning, Maartje J; Hoover, Robert N; Hopper, John L; Hou, Ming-Feng; Hsiung, Chia-Ni; Huang, Guanmengqian; Humphreys, Keith; Ishiguro, Junko; Ito, Hidemi; Iwasaki, Motoki; Iwata, Hiroji; Jakubowska, Anna; Janni, Wolfgang; John, Esther M; Johnson, Nichola; Jones, Kristine; Jones, Michael; Jukkola-Vuorinen, Arja; Kaaks, Rudolf; Kabisch, Maria; Kaczmarek, Katarzyna; Kang, Daehee; Kasuga, Yoshio; Kerin, Michael J; Khan, Sofia; Khusnutdinova, Elza; Kiiski, Johanna I; Kim, Sung-Won; Knight, Julia A; Kosma, Veli-Matti; Kristensen, Vessela N; Krüger, Ute; Kwong, Ava; Lambrechts, Diether; Le Marchand, Loic; Lee, Eunjung; Lee, Min Hyuk; Lee, Jong Won; Neng Lee, Chuen; Lejbkowicz, Flavio; Li, Jingmei; Lilyquist, Jenna; Lindblom, Annika; Lissowska, Jolanta; Lo, Wing-Yee; Loibl, Sibylle; Long, Jirong; Lophatananon, Artitaya; Lubinski, Jan; Luccarini, Craig; Lux, Michael P; Ma, Edmond S K; MacInnis, Robert J; Maishman, Tom; Makalic, Enes; Malone, Kathleen E; Kostovska, Ivana Maleva; Mannermaa, Arto; Manoukian, Siranoush; Manson, JoAnn E; Margolin, Sara; Mariapun, Shivaani; Martinez, Maria Elena; Matsuo, Keitaro; Mavroudis, Dimitrios; McKay, James; McLean, Catriona; Meijers-Heijboer, Hanne; Meindl, Alfons; Menéndez, Primitiva; Menon, Usha; Meyer, Jeffery; Miao, Hui; Miller, Nicola; Taib, Nur Aishah Mohd; Muir, Kenneth; Mulligan, Anna Marie; Mulot, Claire; Neuhausen, Susan L; Nevanlinna, Heli; Neven, Patrick; Nielsen, Sune F; Noh, Dong-Young; Nordestgaard, Børge G; Norman, Aaron; Olopade, Olufunmilayo I; Olson, Janet E; Olsson, Håkan; Olswold, Curtis; Orr, Nick; Pankratz, V Shane; Park, Sue K; Park-Simon, Tjoung-Won; Lloyd, Rachel; Perez, Jose I A; Peterlongo, Paolo; Peto, Julian; Phillips, Kelly-Anne; Pinchev, Mila; Plaseska-Karanfilska, Dijana; Prentice, Ross; Presneau, Nadege; Prokofyeva, Darya; Pugh, Elizabeth; Pylkäs, Katri; Rack, Brigitte; Radice, Paolo; Rahman, Nazneen; Rennert, Gadi; Rennert, Hedy S; Rhenius, Valerie; Romero, Atocha; Romm, Jane; Ruddy, Kathryn J; Rüdiger, Thomas; Rudolph, Anja; Ruebner, Matthias; Rutgers, Emiel J T; Saloustros, Emmanouil; Sandler, Dale P; Sangrajrang, Suleeporn; Sawyer, Elinor J; Schmidt, Daniel F; Schmutzler, Rita K; Schneeweiss, Andreas; Schoemaker, Minouk J; Schumacher, Fredrick; Schürmann, Peter; Scott, Rodney J; Scott, Christopher; Seal, Sheila; Seynaeve, Caroline; Shah, Mitul; Sharma, Priyanka; Shen, Chen-Yang; Sheng, Grace; Sherman, Mark E; Shrubsole, Martha J; Shu, Xiao-Ou; Smeets, Ann; Sohn, Christof; Southey, Melissa C; Spinelli, John J; Stegmaier, Christa; Stewart-Brown, Sarah; Stone, Jennifer; Stram, Daniel O; Surowy, Harald; Swerdlow, Anthony; Tamimi, Rulla; Taylor, Jack A; Tengström, Maria; Teo, Soo H; Beth Terry, Mary; Tessier, Daniel C; Thanasitthichai, Somchai; Thöne, Kathrin; Tollenaar, Rob A E M; Tomlinson, Ian; Tong, Ling; Torres, Diana; Truong, Thérèse; Tseng, Chiu-Chen; Tsugane, Shoichiro; Ulmer, Hans-Ulrich; Ursin, Giske; Untch, Michael; Vachon, Celine; van Asperen, Christi J; Van Den Berg, David; van den Ouweland, Ans M W; van der Kolk, Lizet; van der Luijt, Rob B; Vincent, Daniel; Vollenweider, Jason; Waisfisz, Quinten; Wang-Gohrke, Shan; Weinberg, Clarice R; Wendt, Camilla; Whittemore, Alice S; Wildiers, Hans; Willett, Walter; Winqvist, Robert; Wolk, Alicja; Wu, Anna H; Xia, Lucy; Yamaji, Taiki; Yang, Xiaohong R; Har Yip, Cheng; Yoo, Keun-Young; Yu, Jyh-Cherng; Zheng, Wei; Zheng, Ying; Zhu, Bin; Ziogas, Argyrios; Ziv, Elad; Lakhani, Sunil R; Antoniou, Antonis C; Droit, Arnaud; Andrulis, Irene L; Amos, Christopher I; Couch, Fergus J; Pharoah, Paul D P; Chang-Claude, Jenny; Hall, Per; Hunter, David J; Milne, Roger L; García-Closas, Montserrat; Schmidt, Marjanka K; Chanock, Stephen J; Dunning, Alison M; Edwards, Stacey L; Bader, Gary D; Chenevix-Trench, Georgia; Simard, Jacques; Kraft, Peter; Easton, Douglas F

    2017-11-02

    Breast cancer risk is influenced by rare coding variants in susceptibility genes, such as BRCA1, and many common, mostly non-coding variants. However, much of the genetic contribution to breast cancer risk remains unknown. Here we report the results of a genome-wide association study of breast cancer in 122,977 cases and 105,974 controls of European ancestry and 14,068 cases and 13,104 controls of East Asian ancestry. We identified 65 new loci that are associated with overall breast cancer risk at P < 5 × 10 -8 . The majority of credible risk single-nucleotide polymorphisms in these loci fall in distal regulatory elements, and by integrating in silico data to predict target genes in breast cells at each locus, we demonstrate a strong overlap between candidate target genes and somatic driver genes in breast tumours. We also find that heritability of breast cancer due to all single-nucleotide polymorphisms in regulatory features was 2-5-fold enriched relative to the genome-wide average, with strong enrichment for particular transcription factor binding sites. These results provide further insight into genetic susceptibility to breast cancer and will improve the use of genetic risk scores for individualized screening and prevention.

  2. 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

  3. Potential of isotope analysis (C, Cl) to identify dechlorination mechanisms

    Science.gov (United States)

    Cretnik, Stefan; Thoreson, Kristen; Bernstein, Anat; Ebert, Karin; Buchner, Daniel; Laskov, Christine; Haderlein, Stefan; Shouakar-Stash, Orfan; Kliegman, Sarah; McNeill, Kristopher; Elsner, Martin

    2013-04-01

    Chloroethenes are commonly used in industrial applications, and detected as carcinogenic contaminants in the environment. Their dehalogenation is of environmental importance in remediation processes. However, a detailed understanding frequently accounted problem is the accumulation of toxic degradation products such as cis-dichloroethylene (cis-DCE) at contaminated sites. Several studies have addressed the reductive dehalogenation reactions using biotic and abiotic model systems, but a crucial question in this context has remained open: Do environmental transformations occur by the same mechanism as in their corresponding in vitro model systems? The presented study shows the potential to close this research gap using the latest developments in compound specific chlorine isotope analysis, which make it possible to routinely measure chlorine isotope fractionation of chloroethenes in environmental samples and complex reaction mixtures.1,2 In particular, such chlorine isotope analysis enables the measurement of isotope fractionation for two elements (i.e., C and Cl) in chloroethenes. When isotope values of both elements are plotted against each other, different slopes reflect different underlying mechanisms and are remarkably insensitive towards masking. Our results suggest that different microbial strains (G. lovleyi strain SZ, D. hafniense Y51) and the isolated cofactor cobalamin employ similar mechanisms of reductive dechlorination of TCE. In contrast, evidence for a different mechanism was obtained with cobaloxime cautioning its use as a model for biodegradation. The study shows the potential of the dual isotope approach as a tool to directly compare transformation mechanisms of environmental scenarios, biotic transformations, and their putative chemical lab scale systems. Furthermore, it serves as an essential reference when using the dual isotope approach to assess the fate of chlorinated compounds in the environment.

  4. 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

  5. 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)

  6. 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...

  7. 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

  8. Maximum covariance analysis to identify intraseasonal oscillations over tropical Brazil

    Science.gov (United States)

    Barreto, Naurinete J. C.; Mesquita, Michel d. S.; Mendes, David; Spyrides, Maria H. C.; Pedra, George U.; Lucio, Paulo S.

    2017-09-01

    A reliable prognosis of extreme precipitation events in the tropics is arguably challenging to obtain due to the interaction of meteorological systems at various time scales. A pivotal component of the global climate variability is the so-called intraseasonal oscillations, phenomena that occur between 20 and 100 days. The Madden-Julian Oscillation (MJO), which is directly related to the modulation of convective precipitation in the equatorial belt, is considered the primary oscillation in the tropical region. The aim of this study is to diagnose the connection between the MJO signal and the regional intraseasonal rainfall variability over tropical Brazil. This is achieved through the development of an index called Multivariate Intraseasonal Index for Tropical Brazil (MITB). This index is based on Maximum Covariance Analysis (MCA) applied to the filtered daily anomalies of rainfall data over tropical Brazil against a group of covariates consisting of: outgoing longwave radiation and the zonal component u of the wind at 850 and 200 hPa. The first two MCA modes, which were used to create the { MITB}_1 and { MITB}_2 indices, represent 65 and 16 % of the explained variance, respectively. The combined multivariate index was able to satisfactorily represent the pattern of intraseasonal variability over tropical Brazil, showing that there are periods of activation and inhibition of precipitation connected with the pattern of MJO propagation. The MITB index could potentially be used as a diagnostic tool for intraseasonal forecasting.

  9. Identifying a preservation zone using multicriteria decision analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farashi, A.; Naderi, M.; Parvian, N.

    2016-07-01

    Zoning of a protected area is an approach to partition landscape into various land use units. The management of these landscape units can reduce conflicts caused by human activities. Tandoreh National Park is one of the most biologically diverse, protected areas in Iran. Although the area is generally designed to protect biodiversity, there are many conflicts between biodiversity conservation and human activities. For instance, the area is highly controversial and has been considered as an impediment to local economic development, such as tourism, grazing, road construction, and cultivation. In order to reduce human conflicts with biodiversity conservation in Tandoreh National Park, safe zones need to be established and human activities need to be moved out of the zones. In this study we used a systematic methodology to integrate a participatory process with Geographic Information Systems (GIS) using a multi–criteria decision analysis (MCDA) technique to guide a zoning scheme for the Tandoreh National Park, Iran. Our results show that the northern and eastern parts of the Tandoreh National Park that were close to rural areas and farmlands returned less desirability for selection as a preservation area. Rocky Mountains were the most important and most destructed areas and abandoned plains were the least important criteria for preservation in the area. Furthermore, the results reveal that the land properties were considered to be important for protection based on the obtaine. (Author)

  10. 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

  11. CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

    International Nuclear Information System (INIS)

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

    2011-01-01

    The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

  12. 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 ...

  13. 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.

  14. Meconium microbiome analysis identifies bacteria correlated with premature birth.

    Directory of Open Access Journals (Sweden)

    Alexandria N Ardissone

    Full Text Available Preterm birth is the second leading cause of death in children under the age of five years worldwide, but the etiology of many cases remains enigmatic. The dogma that the fetus resides in a sterile environment is being challenged by recent findings and the question has arisen whether microbes that colonize the fetus may be related to preterm birth. It has been posited that meconium reflects the in-utero microbial environment. In this study, correlations between fetal intestinal bacteria from meconium and gestational age were examined in order to suggest underlying mechanisms that may contribute to preterm birth.Meconium from 52 infants ranging in gestational age from 23 to 41 weeks was collected, the DNA extracted, and 16S rRNA analysis performed. Resulting taxa of microbes were correlated to clinical variables and also compared to previous studies of amniotic fluid and other human microbiome niches.Increased detection of bacterial 16S rRNA in meconium of infants of <33 weeks gestational age was observed. Approximately 61·1% of reads sequenced were classified to genera that have been reported in amniotic fluid. Gestational age had the largest influence on microbial community structure (R = 0·161; p = 0·029, while mode of delivery (C-section versus vaginal delivery had an effect as well (R = 0·100; p = 0·044. Enterobacter, Enterococcus, Lactobacillus, Photorhabdus, and Tannerella, were negatively correlated with gestational age and have been reported to incite inflammatory responses, suggesting a causative role in premature birth.This provides the first evidence to support the hypothesis that the fetal intestinal microbiome derived from swallowed amniotic fluid may be involved in the inflammatory response that leads to premature birth.

  15. 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.

  16. 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

  17. 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

  18. 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)

  19. CoreFlow: A computational platform for integration, analysis and modeling of complex biological data

    DEFF Research Database (Denmark)

    Pasculescu, Adrian; Schoof, Erwin; Creixell, Pau

    2014-01-01

    between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion......A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which...... provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts...

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. Quantitative Proteomic Analysis of Optimal Cutting Temperature (OCT) Embedded Core-Needle Biopsy of Lung Cancer

    Science.gov (United States)

    Zhao, Xiaozheng; Huffman, Kenneth E.; Fujimoto, Junya; Canales, Jamie Rodriguez; Girard, Luc; Nie, Guangjun; Heymach, John V.; Wistuba, Igacio I.; Minna, John D.; Yu, Yonghao

    2017-10-01

    With recent advances in understanding the genomic underpinnings and oncogenic drivers of pathogenesis in different subtypes, it is increasingly clear that proper pretreatment diagnostics are essential for the choice of appropriate treatment options for non-small cell lung cancer (NSCLC). Tumor tissue preservation in optimal cutting temperature (OCT) compound is commonly used in the surgical suite. However, proteins recovered from OCT-embedded specimens pose a challenge for LC-MS/MS experiments, due to the large amounts of polymers present in OCT. Here we present a simple workflow for whole proteome analysis of OCT-embedded NSCLC tissue samples, which involves a simple trichloroacetic acid precipitation step. Comparisons of protein recovery between frozen versus OCT-embedded tissue showed excellent consistency with more than 9200 proteins identified. Using an isobaric labeling strategy, we quantified more than 5400 proteins in tumor versus normal OCT-embedded core needle biopsy samples. Gene ontology analysis indicated that a number of proliferative as well as squamous cell carcinoma (SqCC) marker proteins were overexpressed in the tumor, consistent with the patient's pathology based diagnosis of "poorly differentiated SqCC". Among the most downregulated proteins in the tumor sample, we noted a number of proteins with potential immunomodulatory functions. Finally, interrogation of the aberrantly expressed proteins using a candidate approach and cross-referencing with publicly available databases led to the identification of potential druggable targets in DNA replication and DNA damage repair pathways. We conclude that our approach allows LC-MS/MS proteomic analyses on OCT-embedded lung cancer specimens, opening the way to bring powerful proteomics into the clinic. [Figure not available: see fulltext.

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. Core group approach to identify college students at risk for sexually transmitted infections "Core group" para identificar universitários em risco para infecções sexualmente transmissíveis

    Directory of Open Access Journals (Sweden)

    Miguel A Sánchez-Alemán

    2008-06-01

    Full Text Available OBJECTIVE: To analyze the core group for sexually transmitted infections (STI among college students. METHODS: Cross-sectional study carried out in a convenience sample comprising 711 college students of the public university of Morelos, Mexico, between 2001 and 2003. Sociodemographic and sexual behavior information were collected using self-applied questionnaires. Herpes simplex 2 (HSV-2 infection was tested in the blood. The number of sexual partners in the last year and cocaine consumption were used as indicators to construct the dependent variable "level of STI risk" in three categories: low, medium and high risk (core group. A multinomial analysis was conducted to evaluate whether different sex behaviors were associated with the variable "level of STI risk". RESULTS: There was significant association between HSV-2 seroprevalence and the variable "level of STI risk": 13%, 5.6% and 3.8% were found in high (core group, medium and low categories, respectively. There were gender differences regarding the core group. Men started having sexual intercourse earlier, had more sex partners, higher alcohol and drug consumption, higher frequency of sex intercourse with sex workers, exchanging sex for money, occasional and concurrent partners compared to women. CONCLUSIONS: The study findings suggest existing contextual characteristics in the study population that affect their sex behavior. In Mexico, the cultural conception of sexuality is determined mainly by gender differences where men engage in higher risky sexual behavior than women.OBJETIVO: Identificar al grupo core de infecciones de transmisión sexual (ITS en una población de estudiantes universitarios mexicanos. MÉTODOS: Se realizó un estudio transversal en una muestra por conveniencia que incluyó 711 estudiantes de una universidad pública de Morelos, México, entre 2001 y 2003. Las características sociodemográficas y de comportamiento sexual se obtuvieron mediante un cuestionario auto

  14. 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.

  15. 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.)

  16. 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

  17. Suppression subtractive hybridization and comparative expression analysis to identify developmentally regulated genes in filamentous fungi.

    Science.gov (United States)

    Gesing, Stefan; Schindler, Daniel; Nowrousian, Minou

    2013-09-01

    Ascomycetes differentiate four major morphological types of fruiting bodies (apothecia, perithecia, pseudothecia and cleistothecia) that are derived from an ancestral fruiting body. Thus, fruiting body differentiation is most likely controlled by a set of common core genes. One way to identify such genes is to search for genes with evolutionary conserved expression patterns. Using suppression subtractive hybridization (SSH), we selected differentially expressed transcripts in Pyronema confluens (Pezizales) by comparing two cDNA libraries specific for sexual and for vegetative development, respectively. The expression patterns of selected genes from both libraries were verified by quantitative real time PCR. Expression of several corresponding homologous genes was found to be conserved in two members of the Sordariales (Sordaria macrospora and Neurospora crassa), a derived group of ascomycetes that is only distantly related to the Pezizales. Knockout studies with N. crassa orthologues of differentially regulated genes revealed a functional role during fruiting body development for the gene NCU05079, encoding a putative MFS peptide transporter. These data indicate conserved gene expression patterns and a functional role of the corresponding genes during fruiting body development; such genes are candidates of choice for further functional analysis. © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. Competing endogenous RNA network analysis identifies critical genes among the different breast cancer subtypes.

    Science.gov (United States)

    Chen, Juan; Xu, Juan; Li, Yongsheng; Zhang, Jinwen; Chen, Hong; Lu, Jianping; Wang, Zishan; Zhao, Xueying; Xu, Kang; Li, Yixue; Li, Xia; Zhang, Yan

    2017-02-07

    Although competing endogenous RNAs (ceRNAs) have been implicated in many solid tumors, their roles in breast cancer subtypes are not well understood. We therefore generated a ceRNA network for each subtype based on the significance of both, positive co-expression and the shared miRNAs, in the corresponding subtype miRNA dys-regulatory network, which was constructed based on negative regulations between differentially expressed miRNAs and targets. All four subtype ceRNA networks exhibited scale-free architecture and showed that the common ceRNAs were at the core of the networks. Furthermore, the common ceRNA hubs had greater connectivity than the subtype-specific hubs. Functional analysis of the common subtype ceRNA hubs highlighted factors involved in proliferation, MAPK signaling pathways and tube morphogenesis. Subtype-specific ceRNA hubs highlighted unique subtype-specific pathways, like the estrogen response and inflammatory pathways in the luminal subtypes or the factors involved in the coagulation process that participates in the basal-like subtype. Ultimately, we identified 29 critical subtype-specific ceRNA hubs that characterized the different breast cancer subtypes. Our study thus provides new insight into the common and specific subtype ceRNA interactions that define the different categories of breast cancer and enhances our understanding of the pathology underlying the different breast cancer subtypes, which can have prognostic and therapeutic implications in the future.

  19. Site-Mutation of Hydrophobic Core Residues Synchronically Poise Super Interleukin 2 for Signaling: Identifying Distant Structural Effects through Affordable Computations

    Directory of Open Access Journals (Sweden)

    Longcan Mei

    2018-03-01

    Full Text Available A superkine variant of interleukin-2 with six site mutations away from the binding interface developed from the yeast display technique has been previously characterized as undergoing a distal structure alteration which is responsible for its super-potency and provides an elegant case study with which to get insight about how to utilize allosteric effect to achieve desirable protein functions. By examining the dynamic network and the allosteric pathways related to those mutated residues using various computational approaches, we found that nanosecond time scale all-atom molecular dynamics simulations can identify the dynamic network as efficient as an ensemble algorithm. The differentiated pathways for the six core residues form a dynamic network that outlines the area of structure alteration. The results offer potentials of using affordable computing power to predict allosteric structure of mutants in knowledge-based mutagenesis.

  20. 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

  1. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  2. Post-clip placement MRI following second-look US-guided core biopsy for suspicious lesions identified on breast MRI

    International Nuclear Information System (INIS)

    Song, Sung Eun; Cho, Nariya; Han, Wonshik

    2017-01-01

    To evaluate whether the post-clip placement MRI following second-look ultrasound (US)-guided core biopsy is useful to confirm the adequate sampling of suspicious lesions identified on breast MRI. Between 2014 and 2016, 31 consecutive women with 34 suspicious lesions that had not been identified on previous mammography or US were detected using MRI. Among them, 26 women with 29 lesions (mean size 1.5 cm, range 0.5-5.8 cm) found by second-look US underwent US-guided biopsy, subsequent clip insertion and post-clip placement MRI. Five women with five lesions that were not found by second-look US underwent MRI-guided biopsy. The technical success rate and lesion characteristics were described. The technical success rate was 96.6% (28/29). One failure case was a benign, 1.1-cm non-mass enhancement. Of the 28 success cases, 23 (82.1%) were masses and 5 (17.9%) were non-mass enhancements; 17 (60.7%) were benign, 4 (14.3%) were high-risk and 7 (25.0%) were malignant lesions. The technical success rate was 100% (28/28) for masses and 83.3% (5/6) for non-mass enhancements. Post-clip placement MRI following US-guided biopsy is useful in confirming the adequate sampling of lesions identified on MRI. This method could be an alternative to MRI-guided biopsy for lesions visible on US. (orig.)

  3. Post-clip placement MRI following second-look US-guided core biopsy for suspicious lesions identified on breast MRI

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Eun [Seoul National University Hospital, Department of Radiology, Seoul (Korea, Republic of); Korea University Anam Hospital, Korea University College of Medicine, Department of Radiology, Seoul (Korea, Republic of); Cho, Nariya [Seoul National University Hospital, Department of Radiology, Seoul (Korea, Republic of); Seoul National University College of Medicine, Department of Radiology, Seoul (Korea, Republic of); Seoul National University Medical Research Centre, Institute of Radiation Medicine, Seoul (Korea, Republic of); Han, Wonshik [Seoul National University Hospital, Department of Surgery, Seoul (Korea, Republic of)

    2017-12-15

    To evaluate whether the post-clip placement MRI following second-look ultrasound (US)-guided core biopsy is useful to confirm the adequate sampling of suspicious lesions identified on breast MRI. Between 2014 and 2016, 31 consecutive women with 34 suspicious lesions that had not been identified on previous mammography or US were detected using MRI. Among them, 26 women with 29 lesions (mean size 1.5 cm, range 0.5-5.8 cm) found by second-look US underwent US-guided biopsy, subsequent clip insertion and post-clip placement MRI. Five women with five lesions that were not found by second-look US underwent MRI-guided biopsy. The technical success rate and lesion characteristics were described. The technical success rate was 96.6% (28/29). One failure case was a benign, 1.1-cm non-mass enhancement. Of the 28 success cases, 23 (82.1%) were masses and 5 (17.9%) were non-mass enhancements; 17 (60.7%) were benign, 4 (14.3%) were high-risk and 7 (25.0%) were malignant lesions. The technical success rate was 100% (28/28) for masses and 83.3% (5/6) for non-mass enhancements. Post-clip placement MRI following US-guided biopsy is useful in confirming the adequate sampling of lesions identified on MRI. This method could be an alternative to MRI-guided biopsy for lesions visible on US. (orig.)

  4. 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

  5. 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

  6. 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

  7. The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

    2006-07-15

    This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

  8. Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

    International Nuclear Information System (INIS)

    Pérez-Valseca, A.-D.; Espinosa-Paredes, G.; François, J.L.; Vázquez Rodríguez, A.; Martín-del-Campo, C.

    2017-01-01

    Highlights: • Methodology based on Monte Carlo simulation. • Sensitivity analysis of Lead Fast Reactor (LFR). • Uncertainty and regression analysis of LFR. • 10% change in the core inlet flow, the response in thermal power change is 0.58%. • 2.5% change in the inlet lead temperature the response is 1.87% in power. - Abstract: The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

  9. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  10. CoreFlow: a computational platform for integration, analysis and modeling of complex biological data.

    Science.gov (United States)

    Pasculescu, Adrian; Schoof, Erwin M; Creixell, Pau; Zheng, Yong; Olhovsky, Marina; Tian, Ruijun; So, Jonathan; Vanderlaan, Rachel D; Pawson, Tony; Linding, Rune; Colwill, Karen

    2014-04-04

    A major challenge in mass spectrometry and other large-scale applications is how to handle, integrate, and model the data that is produced. Given the speed at which technology advances and the need to keep pace with biological experiments, we designed a computational platform, CoreFlow, which provides programmers with a framework to manage data in real-time. It allows users to upload data into a relational database (MySQL), and to create custom scripts in high-level languages such as R, Python, or Perl for processing, correcting and modeling this data. CoreFlow organizes these scripts into project-specific pipelines, tracks interdependencies between related tasks, and enables the generation of summary reports as well as publication-quality images. As a result, the gap between experimental and computational components of a typical large-scale biology project is reduced, decreasing the time between data generation, analysis and manuscript writing. CoreFlow is being released to the scientific community as an open-sourced software package complete with proteomics-specific examples, which include corrections for incomplete isotopic labeling of peptides (SILAC) or arginine-to-proline conversion, and modeling of multiple/selected reaction monitoring (MRM/SRM) results. CoreFlow was purposely designed as an environment for programmers to rapidly perform data analysis. These analyses are assembled into project-specific workflows that are readily shared with biologists to guide the next stages of experimentation. Its simple yet powerful interface provides a structure where scripts can be written and tested virtually simultaneously to shorten the life cycle of code development for a particular task. The scripts are exposed at every step so that a user can quickly see the relationships between the data, the assumptions that have been made, and the manipulations that have been performed. Since the scripts use commonly available programming languages, they can easily be

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

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

    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

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

    International Nuclear Information System (INIS)

    RASMUSSEN, J.H.

    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-AZ-102

  15. Analysis of sodium-void experiments in ZPPR-3 modified Phase 3 core

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, T.

    1978-08-01

    An analysis is presented of a series of sodium-void reactivity measurements performed in assembly 3 of Zero Power Plutonium Reactor (ZPPR-3), a mockup of the US Demoplant. In this series, large-zone sodium-void effects were studied in detail in the presence of many singularities, namely, control rods (CRs) and control rod positions (CRPs). The Karlsruhe data-and-method have been applied to an analysis of these experiments, and the results are presented. The work is aimed at complementing the sodium-void reactivity analysis based on the SNEAK experiments, where it was difficult to simulate a large plutonium-core of a prototype fast breeder reactor.

  16. 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)

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

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da.

    1991-01-01

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

  18. 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)

  19. 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

  20. 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.

  1. 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

  2. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  3. Tank 241-TX-113 rotary mode core sampling and analysis plan

    International Nuclear Information System (INIS)

    McCain, D.J.

    1998-01-01

    This sampling and analysis plan (SAP) identities characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for push mode core samples from tank 241-TX-113 (TX-113). The Tank Characterization Technical Sampling Basis document identities Retrieval, Pretreatment and Immobilization as an issue that applies to tank TX-113. As a result, a 150 gram composite of solids shall be made and archived for that program. This tank is not on a Watch List

  4. 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...

  5. Numerical analysis of sandwich beam with corrugated core under three-point bending

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbeck, Leszek [Poznan University of Technology, Institute of Mathematics Piotrowo Street No. 5, 60-965 Poznan (Poland); Grygorowicz, Magdalena; Paczos, Piotr [Poznan University of Technology, Institute of Applied Mechanics Jana Pawla IIStreet No. 24, 60-965 Poznan (Poland)

    2015-03-10

    The strength problem of sandwich beam with corrugated core under three-point bending is presented.The beam are made of steel and formed by three mutually orthogonal corrugated layers. The finite element analysis (FEA) of the sandwich beam is performed with the use of the FEM system - ABAQUS. The relationship between the applied load and deflection in three-point bending is considered.

  6. Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire

    Science.gov (United States)

    2016-11-01

    Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire E n g in e e r R e s e a rc h a n d...id, age of the concrete being evaluated and tests performed...4 3 Preface This study was conducted in support of the Air Force Civil Engineer Center (AFCEC) to assess concrete obtained from Pease

  7. Analysis on High Temperature Aging Property of Self-brazing Aluminum Honeycomb Core at Middle Temperature

    Directory of Open Access Journals (Sweden)

    ZHAO Huan

    2016-11-01

    Full Text Available Tension-shear test was carried out on middle temperature self-brazing aluminum honeycomb cores after high temperature aging by micro mechanical test system, and the microstructure and component of the joints were observed and analyzed using scanning electron microscopy and energy dispersive spectroscopy to study the relationship between brazing seam microstructure, component and high temperature aging properties. Results show that the tensile-shear strength of aluminum honeycomb core joints brazed by 1060 aluminum foil and aluminum composite brazing plate after high temperature aging(200℃/12h, 200℃/24h, 200℃/36h is similar to that of as-welded joints, and the weak part of the joint is the base metal which is near the brazing joint. The observation and analysis of the aluminum honeycomb core microstructure and component show that the component of Zn, Sn at brazing seam is not much affected and no compound phase formed after high temperature aging; therefore, the main reason for good high temperature aging performance of self-brazing aluminum honeycomb core is that no obvious change of brazing seam microstructure and component occurs.

  8. Provenance of whitefish in the Gulf of Bothnia determined by elemental analysis of otolith cores

    Science.gov (United States)

    Lill, J.-O.; Finnäs, V.; Slotte, J. M. K.; Jokikokko, E.; Heimbrand, Y.; Hägerstrand, H.

    2018-02-01

    The strontium concentration in the core of otoliths was used to determine the provenance of whitefish found in the Gulf of Bothnia, Baltic Sea. To that end, a database of strontium concentration in fish otoliths representing different habitats (sea, river and fresh water) had to be built. Otoliths from juvenile whitefish were therefore collected from freshwater ponds at 5 hatcheries, from adult whitefish from 6 spawning sites at sea along the Finnish west coast, and from adult whitefish ascending to spawn in the Torne River, in total 67 otoliths. PIXE was applied to determine the elemental concentrations in these otoliths. While otoliths from the juveniles raised in the freshwater ponds showed low but varying strontium concentrations (194-1664 μg/g,), otoliths from sea-spawning fish showed high uniform strontium levels (3720-4333 μg/g). The otolith core analysis of whitefish from Torne River showed large variations in the strontium concentrations (1525-3650 μg/g). These otolith data form a database to be used for provenance studies of wild adult whitefish caught at sea. The applicability of the database was evaluated by analyzing the core of polished otoliths from 11 whitefish from a test site at sea in the Larsmo archipelago. Our results show that by analyzing strontium in the otolith core, we can differentiate between hatchery-origin and wild-origin whitefish, but not always between river and sea spawning whitefish.

  9. 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)

  10. 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)

  11. Analysis of forces on core structures during a loss-of-coolant accident. Final report

    International Nuclear Information System (INIS)

    Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

    1980-08-01

    There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

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

    International Nuclear Information System (INIS)

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

    1990-04-01

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

  13. Investigation on macroscopic cross section model for BWR pin-by-pin core analysis - 118

    International Nuclear Information System (INIS)

    Fujita, T.; Tada, K.; Yamamoto, A.; Yamane, Y.; Kosaka, S.; Hirano, G.

    2010-01-01

    A cross section model used in the pin-by-pin core analysis for BWR is investigated. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of state and history variables that have influences on the cross section and are tabulated prior to the core calculations. Variation of a cross section in a core simulator is classified into two different types, i.e., the instantaneous effect and the history effect. The instantaneous effect is incorporated by the variation of cross section which is caused by the instantaneous change of state variables. For this effect, the exposure, the void fraction, the fuel temperature, the moderator temperature and the control rod are used as indexes. The history effect is the cumulative effect of state variables. We treat this effect with a unified approach using the spectral history. To confirm accuracy of the cross section model, the pin-by-pin fission rate distribution and the k-infinity of fuel assembly which are obtained with the tabulated and the reference cross sections are compared. For the instantaneous effect, the present cross section model well reproduces the reference results for all off-nominal conditions. For the history effect, however, considerable differences both on the pin-by-pin fission rate distribution and the k-infinity are observed at high exposure points. (authors)

  14. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-01-01

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal's mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs

  15. 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

  16. 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.

  17. INSIGHT: an integrated scoping analysis tool for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Noda, Hidefumi; Ito, Nobuaki; Maruyama, Taiji.

    1997-01-01

    An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis. (author)

  18. 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.

  19. 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.

  20. 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

  1. The accuracy of frozen section analysis in ultrasound- guided core needle biopsy of breast lesions

    International Nuclear Information System (INIS)

    Brunner, Andreas H; Sagmeister, Thomas; Kremer, Jolanta; Riss, Paul; Brustmann, Hermann

    2009-01-01

    Limited data are available to evaluate the accuracy of frozen section analysis and ultrasound- guided core needle biopsy of the breast. In a retrospective analysis data of 120 consecutive handheldultrasound- guided 14- gauge automated core needle biopsies (CNB) in 109 consecutive patients with breast lesions between 2006 and 2007 were evaluated. In our outpatient clinic120 CNB were performed. In 59/120 (49.2%) cases we compared histological diagnosis on frozen sections with those on paraffin sections of CNB and finally with the result of open biopsy. Of the cases 42/59 (71.2%) were proved to be malignant and 17/59 (28.8%) to be benign in the definitive histology. 2/59 (3.3%) biopsies had a false negative frozen section result. No false positive results of the intraoperative frozen section analysis were obtained, resulting in a sensitivity, specificity and positive predicting value (PPV) and negative predicting value (NPV) of 95%, 100%, 100% and 90%, respectively. Histological and morphobiological parameters did not show up relevance for correct frozen section analysis. In cases of malignancy time between diagnosis and definitive treatment could not be reduced due to frozen section analysis. The frozen section analysis of suspect breast lesions performed by CNB displays good sensitivity/specificity characteristics. Immediate investigations of CNB is an accurate diagnostic tool and an important step in reducing psychological strain by minimizing the period of uncertainty in patients with breast tumor

  2. 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.

  3. False-negative results of breast core needle biopsies – retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Boba, Marek; Kołtun, Urszula; Bobek-Billewicz, Barbara; Chmielik, Ewa; Eksner, Bartosz; Olejnik, Tomasz

    2011-01-01

    Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. 988 core needle breast biopsies were performed at the Maria Skłodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration

  4. False-negative results of breast core needle biopsies - retrospective analysis of 988 biopsies

    International Nuclear Information System (INIS)

    Boba, M.; Koltun, U.; Bobek-Billewicz, B.; Eksner, B.; Olejnik, T.; Chmielik, E.

    2011-01-01

    Background: Breast cancer is the most common malignant neoplasm and the most common cause of death among women. The core needle biopsy is becoming a universal practice in diagnosing breast lesions suspected of malignancy. Unfortunately, breast core needle biopsies also bear the risk of having false-negative results. Material/Methods: 988 core needle breast biopsies were performed at the Maria Sklodowska-Curie Memorial Cancer Center and Institute of Oncology, Gliwice Branch, between 01 March 2006 and 29 February 2008. Malignant lesions were diagnosed in 426/988 (43.12%) cases, atypical hyperplasia in 69/988 (6.98%), and benign lesions in 493/988 (49.90%) cases. Results: Twenty-two out of 988 biopsies (2.23%) were found to be false negative. Histopathological assessment of tissue specimens was repeated in these cases. In 14/22 (64%) cases, the previous diagnosis of a benign lesion was changed. In 8/22 (36%) cases, the diagnosis of a benign lesion was confirmed. False-negative rate was calculated at 2.2%. The rate of false-negative diagnoses resulting from a radiological mistake was estimated at 36%. The rate of false-negative diagnoses, resulting from histopathological assessment, was 64%. False-negative results caused by a radiological error comprised 1.5% of all histopathologically diagnosed cancers and atypias (sensitivity of 98.5%). There were no false-positive results in our material - the specificity of the method was 100%. Conclusions: Histopathological interpretation is a substantial cause of false-negative results of breast core needle biopsy. Thus, in case of a radiological-histopathological divergence, histopathological analysis of biopsy specimens should be repeated. The main radiological causes of false-negative results of breast core needle biopsy are as follows: sampling from an inappropriate site and histopathological non-homogeneity of cancer infiltration. (authors)

  5. 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.

  6. 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)

  7. 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

  8. Post-clip placement MRI following second-look US-guided core biopsy for suspicious lesions identified on breast MRI.

    Science.gov (United States)

    Song, Sung Eun; Cho, Nariya; Han, Wonshik

    2017-12-01

    To evaluate whether the post-clip placement MRI following second-look ultrasound (US)-guided core biopsy is useful to confirm the adequate sampling of suspicious lesions identified on breast MRI. Between 2014 and 2016, 31 consecutive women with 34 suspicious lesions that had not been identified on previous mammography or US were detected using MRI. Among them, 26 women with 29 lesions (mean size 1.5 cm, range 0.5-5.8 cm) found by second-look US underwent US-guided biopsy, subsequent clip insertion and post-clip placement MRI. Five women with five lesions that were not found by second-look US underwent MRI-guided biopsy. The technical success rate and lesion characteristics were described. The technical success rate was 96.6% (28/29). One failure case was a benign, 1.1-cm non-mass enhancement. Of the 28 success cases, 23 (82.1%) were masses and 5 (17.9%) were non-mass enhancements; 17 (60.7%) were benign, 4 (14.3%) were high-risk and 7 (25.0%) were malignant lesions. The technical success rate was 100% (28/28) for masses and 83.3% (5/6) for non-mass enhancements. Post-clip placement MRI following US-guided biopsy is useful in confirming the adequate sampling of lesions identified on MRI. This method could be an alternative to MRI-guided biopsy for lesions visible on US. • Post-clip MRI is useful for confirming adequate sampling of US-guided biopsy. • Post-clip MRI following US-guided biopsy revealed a 96.6 % technical success rate. • One technical failure case was a benign, 1.1-cm non-mass enhancement. • The technical success rate of US-guided biopsy for non-mass enhancements was 83.3 %.

  9. 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

  10. 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

  11. 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)

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

    Directory of Open Access Journals (Sweden)

    Kostić Miloje

    2015-01-01

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

  13. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  14. 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)

  15. 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)

  16. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

  17. Assessing Reliability of Cellulose Hydrolysis Models to Support Biofuel Process Design – Identifiability and Uncertainty Analysis

    DEFF Research Database (Denmark)

    Sin, Gürkan; Meyer, Anne S.; Gernaey, Krist

    2010-01-01

    The reliability of cellulose hydrolysis models is studied using the NREL model. An identifiability analysis revealed that only 6 out of 26 parameters are identifiable from the available data (typical hydrolysis experiments). Attempting to identify a higher number of parameters (as done in the ori......The reliability of cellulose hydrolysis models is studied using the NREL model. An identifiability analysis revealed that only 6 out of 26 parameters are identifiable from the available data (typical hydrolysis experiments). Attempting to identify a higher number of parameters (as done...

  18. Identifying the source of petroleum pollution in sediment cores of southwest of the Caspian Sea using chemical fingerprinting of aliphatic and alicyclic hydrocarbons.

    Science.gov (United States)

    Shirneshan, Golshan; Bakhtiari, Alireza Riyahi; Memariani, Mahmoud

    2017-02-15

    In this study, the concentration and sources of aliphatic and petroleum markers were investigated in 105 samples of Anzali, Rezvanshahr and Astara cores from the southwest of Caspian Sea. Petroleum importation was diagnosed as a main source in most depths of cores by the results of unresolved complex mixture, carbon preference index and hopanes and steranes. From the chemical diagnostic parameters, petroleum inputs in sediment of cores were determined to be different during years and the sources of hydrocarbons in some sections differed than Anzali and Turkmenistan and Azerbaijan oils. Diagenic ratios in most sediments of upper and middle sections in Astara core were determined to be highly similar to those of Azerbaijan oil, while the presence of Turkmenistan and Anzali oils were detected in a few sections of Anzali and Rezvanshahr cores and only five layers of downer section in Anzali core, respectively. Copyright © 2016. Published by Elsevier Ltd.

  19. Calculation and analysis of burnup and optimum core design in accelerator driven sub-critical system

    International Nuclear Information System (INIS)

    Wang Yuwei; Yang Yongwei; Cui Pengfei

    2011-01-01

    The premise of the accelerator driven sub-critical system (ADS) in the accident is still subcritical, the biggest k eff change with burn time is less than 1.5% and the cladding material, HT9 steel, can withstand the maximum radiation damage, core fuel area is divided into fuel transmutation area and fuel multiplication area, and fuel transmutation area maintains the same fuel composition in the whole process. Through the analysis of the composition of the fuel, shape of core layout and the power distribution, etc., supposed outer and inner Pu enrichment ratio range of 1.0-1.5, then the fuel components of fuel multiplication area was adjusted. Time evolution of k eff was calculated by COUPLED2 which coupled with MCNP and ORIGEN. At the same time the power peaking factors, minoractinides transmutation rate desired to maximization and burnup were considered. A sub-critical system fitting for engineering practice was established. (authors)

  20. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    International Nuclear Information System (INIS)

    Prabhu Gaunkar, N.; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C.; Bulu, I.; Ganesan, K.; Song, Y. Q.

    2015-01-01

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors

  1. SACI - O: A code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Resende Lobo, A.A. de; Soares, P.A.

    1979-03-01

    The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt

  2. 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

  3. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  4. Neutronic analysis of the Three Mile Island Unit 2 ex-core detector response

    International Nuclear Information System (INIS)

    Malloy, D.J.; Chang, Y.I.

    1981-10-01

    A neutronic analysis has been made with respect to the ex-core neutron detector response during the TMI-2 incident. A series of transport theory calculations quantified the impact upon the detector count rate of various core and downcomer conditions. In particular, various combinations of coolant void content and spatial distributions were investigated to yield the resulting transmission of the photoneutron source to the detector. The impact of a hypothetical distributed source within the downcomer region was also examined in order to simulate the potential effect of the release of neutron producing fission products into the coolant. These results are then offered as potential explanations for the anomalous behavior of the detector during the period of approx. 20 minutes through approx. 3 hours following the reactor scram

  5. 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

  6. 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

  7. A Strategy for Identifying Quantitative Trait Genes Using Gene Expression Analysis and Causal Analysis

    Directory of Open Access Journals (Sweden)

    Akira Ishikawa

    2017-11-01

    Full Text Available Large numbers of quantitative trait loci (QTL affecting complex diseases and other quantitative traits have been reported in humans and model animals. However, the genetic architecture of these traits remains elusive due to the difficulty in identifying causal quantitative trait genes (QTGs for common QTL with relatively small phenotypic effects. A traditional strategy based on techniques such as positional cloning does not always enable identification of a single candidate gene for a QTL of interest because it is difficult to narrow down a target genomic interval of the QTL to a very small interval harboring only one gene. A combination of gene expression analysis and statistical causal analysis can greatly reduce the number of candidate genes. This integrated approach provides causal evidence that one of the candidate genes is a putative QTG for the QTL. Using this approach, I have recently succeeded in identifying a single putative QTG for resistance to obesity in mice. Here, I outline the integration approach and discuss its usefulness using my studies as an example.

  8. A Strategy for Identifying Quantitative Trait Genes Using Gene Expression Analysis and Causal Analysis.

    Science.gov (United States)

    Ishikawa, Akira

    2017-11-27

    Large numbers of quantitative trait loci (QTL) affecting complex diseases and other quantitative traits have been reported in humans and model animals. However, the genetic architecture of these traits remains elusive due to the difficulty in identifying causal quantitative trait genes (QTGs) for common QTL with relatively small phenotypic effects. A traditional strategy based on techniques such as positional cloning does not always enable identification of a single candidate gene for a QTL of interest because it is difficult to narrow down a target genomic interval of the QTL to a very small interval harboring only one gene. A combination of gene expression analysis and statistical causal analysis can greatly reduce the number of candidate genes. This integrated approach provides causal evidence that one of the candidate genes is a putative QTG for the QTL. Using this approach, I have recently succeeded in identifying a single putative QTG for resistance to obesity in mice. Here, I outline the integration approach and discuss its usefulness using my studies as an example.

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

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

  11. A finite element thermal analysis of various dowel and core materials

    Directory of Open Access Journals (Sweden)

    Shanti Varghese

    2012-01-01

    Conclusion: Non-metallic dowel and core materials such as fibre reinforced composite dowels (FRC generate greater stress than metallic dowel and core materials. This emphasized the preferable use of the metallic dowel and core materials in the oral environment.

  12. Efficient Culture Adaptation of Hepatitis C Virus Recombinants with Genotype-Specific Core-NS2 by Using Previously Identified Mutations

    DEFF Research Database (Denmark)

    Scheel, Troels Kasper Høyer; Gottwein, Judith M; Carlsen, Thomas H R

    2011-01-01

    Hepatitis C virus (HCV) is an important cause of chronic liver disease, and interferon-based therapy cures only 40 to 80% of patients, depending on HCV genotype. Research was accelerated by genotype 2a (strain JFH1) infectious cell culture systems. We previously developed viable JFH1-based...... (HC-TN and DH6), 1b (DH1 and DH5), and 3a (DBN) isolates, using previously identified adaptive mutations. Introduction of mutations from isolates of the same subtype either led to immediate efficient virus production or accelerated culture adaptation. The DH6 and DH5 recombinants without introduced...... mutations did not adapt to culture. Universal adaptive effects of mutations in NS3 (Q1247L, I1312V, K1398Q, R1408W, and Q1496L) and NS5A (V2418L) were investigated for JFH1-based genotype 1 to 5 core-NS2 recombinants; several mutations conferred adaptation to H77C (1a), J4 (1b), S52 (3a), and SA13 (5a...

  13. Efficacy of the core DNA barcodes in identifying processed and poorly conserved plant materials commonly used in South African traditional medicine

    Directory of Open Access Journals (Sweden)

    Ledile Mankga

    2013-12-01

    Full Text Available Medicinal plants cover a broad range of taxa, which may be phylogenetically less related but morphologically very similar. Such morphological similarity between species may lead to misidentification and inappropriate use. Also the substitution of a medicinal plant by a cheaper alternative (e.g. other non-medicinal plant species, either due to misidentification, or deliberately to cheat consumers, is an issue of growing concern. In this study, we used DNA barcoding to identify commonly used medicinal plants in South Africa. Using the core plant barcodes, matK and rbcLa, obtained from processed and poorly conserved materials sold at the muthi traditional medicine market, we tested efficacy of the barcodes in species discrimination. Based on genetic divergence, PCR amplification efficiency and BLAST algorithm, we revealed varied discriminatory potentials for the DNA barcodes. In general, the barcodes exhibited high discriminatory power, indicating their effectiveness in verifying the identity of the most common plant species traded in South African medicinal markets. BLAST algorithm successfully matched 61% of the queries against a reference database, suggesting that most of the information supplied by sellers at traditional medicinal markets in South Africa is correct. Our findings reinforce the utility of DNA barcoding technique in limiting false identification that can harm public health.

  14. Radiation Therapy Deficiencies Identified During On-Site Dosimetry Visits by the Imaging and Radiation Oncology Core Houston Quality Assurance Center.

    Science.gov (United States)

    Kry, Stephen F; Dromgoole, Lainy; Alvarez, Paola; Leif, Jessica; Molineu, Andrea; Taylor, Paige; Followill, David S

    2017-12-01

    To review the dosimetric, mechanical, and programmatic deficiencies most frequently observed during on-site visits of radiation therapy facilities by the Imaging and Radiation Oncology Core Quality Assurance Center in Houston (IROC Houston). The findings of IROC Houston between 2000 and 2014, including 409 institutions and 1020 linear accelerators (linacs), were compiled. On-site evaluations by IROC Houston include verification of absolute calibration (tolerance of ±3%), relative dosimetric review (tolerances of ±2% between treatment planning system [TPS] calculation and measurement), mechanical evaluation (including multileaf collimator and kilovoltage-megavoltage isocenter evaluation against Task Group [TG]-142 tolerances), and general programmatic review (including institutional quality assurance program vs TG-40 and TG-142). An average of 3.1 deficiencies was identified at each institution visited, a number that has decreased slightly with time. The most common errors are tabulated and include TG-40/TG-142 compliance (82% of institutions were deficient), small field size output factors (59% of institutions had errors ≥3%), and wedge factors (33% of institutions had errors ≥3%). Dosimetric errors of ≥10%, including in beam calibration, were seen at many institutions. There is substantial room for improvement of both dosimetric and programmatic issues in radiation therapy, which should be a high priority for the medical physics community. Particularly relevant was suboptimal beam modeling in the TPS and a corresponding failure to detect these errors by not including TPS data in the linac quality assurance process. Copyright © 2017 Elsevier Inc. All rights reserved.

  15. Analysis of core-periphery organization in protein contact networks reveals groups of structurally and functionally critical residues.

    Science.gov (United States)

    Isaac, Arnold Emerson; Sinha, Sitabhra

    2015-10-01

    The representation of proteins as networks of interacting amino acids, referred to as protein contact networks (PCN), and their subsequent analyses using graph theoretic tools, can provide novel insights into the key functional roles of specific groups of residues. We have characterized the networks corresponding to the native states of 66 proteins (belonging to different families) in terms of their core-periphery organization. The resulting hierarchical classification of the amino acid constituents of a protein arranges the residues into successive layers - having higher core order - with increasing connection density, ranging from a sparsely linked periphery to a densely intra-connected core (distinct from the earlier concept of protein core defined in terms of the three-dimensional geometry of the native state, which has least solvent accessibility). Our results show that residues in the inner cores are more conserved than those at the periphery. Underlining the functional importance of the network core, we see that the receptor sites for known ligand molecules of most proteins occur in the innermost core. Furthermore, the association of residues with structural pockets and cavities in binding or active sites increases with the core order. From mutation sensitivity analysis, we show that the probability of deleterious or intolerant mutations also increases with the core order. We also show that stabilization centre residues are in the innermost cores, suggesting that the network core is critically important in maintaining the structural stability of the protein. A publicly available Web resource for performing core-periphery analysis of any protein whose native state is known has been made available by us at http://www.imsc.res.in/ ~sitabhra/proteinKcore/index.html.

  16. 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

  17. 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)

  18. 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

  19. Modeling analysis of pulsed magnetization process of magnetic core based on inverse Jiles-Atherton model

    Science.gov (United States)

    Liu, Yi; Zhang, He; Liu, Siwei; Lin, Fuchang

    2018-05-01

    The J-A (Jiles-Atherton) model is widely used to describe the magnetization characteristics of magnetic cores in a low-frequency alternating field. However, this model is deficient in the quantitative analysis of the eddy current loss and residual loss in a high-frequency magnetic field. Based on the decomposition of magnetization intensity, an inverse J-A model is established which uses magnetic flux density B as an input variable. Static and dynamic core losses under high frequency excitation are separated based on the inverse J-A model. Optimized parameters of the inverse J-A model are obtained based on particle swarm optimization. The platform for the pulsed magnetization characteristic test is designed and constructed. The hysteresis curves of ferrite and Fe-based nanocrystalline cores at high magnetization rates are measured. The simulated and measured hysteresis curves are presented and compared. It is found that the inverse J-A model can be used to describe the magnetization characteristics at high magnetization rates and to separate the static loss and dynamic loss accurately.

  20. 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.

  1. Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

    1978-07-01

    The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

  2. 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)

  3. 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

  4. 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...

  5. 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