WorldWideScience

Sample records for total core damage

  1. Core damage risk indicators

    International Nuclear Information System (INIS)

    Szikszai, T.

    1994-01-01

    The purpose of this document is to show a method for the fast recalculation of the PSA. To avoid the information loose, it is necessary to simplify the PSA models, or at least reorganize them. The method, introduced in this document, require that preparation, so we try to show, how to do that. This document is an introduction. This is the starting point of the work related to the development of the risk indicators. In the future, with the application of this method, we are going to show an everyday use of the PSA results to produce the indicators of the core damage risk. There are two different indicators of the plant safety performance, related to the core damage risk. The first is the core damage frequency indicator (CDFI), and the second is the core damage probability indicator (CDPI). Of course, we cannot describe all of the possible ways to use these indicators, rather we will try to introduce the requirements to establish such an indicator system and the calculation process

  2. 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)

  3. 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

  4. TMI-2 core damage: a summary of present knowledge

    International Nuclear Information System (INIS)

    Owen, D.E.; Mason, R.E.; Meininger, R.D.; Franz, W.A.

    1983-01-01

    Extensive fuel damage (oxidation and fragmentation) has occurred and the top approx. 1.5 m of the center portion of the TMI-2 core has relocated. The fuel fragmentation extends outward to slightly beyond one-half the core radius in the direction examined by the CCTV camera. While the radial extent of core fragmentation in other directions was not directly observed, control and spider drop data and in-core instrument data suggest that the core void is roughly symmetrical, although there are a few indications of severe fuel damage extending to the core periphery. The core material fragmented into a broad range of particle sizes, extending down to a few microns. APSR movement data, the observation of damaged fuel assemblies hanging unsupported from the bottom of the reactor upper plenum structure, and the observation of once-molten stainless steel immediately above the active core indicate high temperatures (up to at least 1720 K) extended to the very top of the core. The relative lack of damage to the underside of the plenum structure implies a sharp temperature demarcation at the core/plenum interface. Filter debris and leadscrew deposit analyses indicate extensive high temperature core materials interaction, melting of the Ag-In-Cd control material, and transport of particulate control material to the plenum and out of the vessel

  5. 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

  6. Present status and needs of research on severe core damage

    International Nuclear Information System (INIS)

    1982-05-01

    The needs for research on severe core damage accident have been emphasized recently, in particular, since TMI-2 accident. The Severe Core Damage Research Task Force was established by the Divisions of Reactor Safety and Reactor Safety Evaluation to evaluate individual phenomenon, to survey the present status of research and to provide the recommended research subjects on severe accidents. This report describes the accident phenomena involving some analytical results, status of research and recommended research subjects on severe core damage accidents, divided into accident sequence, fuel damage, and molten material behavior, fission product behavior, hydrogen generation and combustion, steam explosion and containment integrity. (author)

  7. An examination of impact damage in glass-phenolic and aluminum honeycomb core composite panels

    Science.gov (United States)

    Nettles, A. T.; Lance, D. G.; Hodge, A. J.

    1990-01-01

    An examination of low velocity impact damage to glass-phenolic and aluminum core honeycomb sandwich panels with carbon-epoxy facesheets is presented. An instrumented drop weight impact test apparatus was utilized to inflict damage at energy ranges between 0.7 and 4.2 joules. Specimens were checked for extent of damage by cross sectional examination. The effect of core damage was assessed by subjecting impact-damaged beams to four-point bend tests. Skin-only specimens (facings not bonded to honeycomb) were also tested for comparison purposes. Results show that core buckling is the first damage mode, followed by delaminations in the facings, matrix cracking, and finally fiber breakage. The aluminum honeycomb panels exhibited a larger core damage zone and more facing delaminations than the glass-phenolic core, but could withstand more shear stress when damaged than the glass-phenolic core specimens.

  8. Characterization of the Fault Core and Damage Zone of the Borrego Fault, 2010 M7.2 Rupture

    Science.gov (United States)

    Dorsey, M. T.; Rockwell, T. K.; Girty, G.; Ostermeijer, G.; Mitchell, T. M.; Fletcher, J. M.

    2017-12-01

    We collected a continuous sample of the fault core and 23 samples of the damage zone out to 52 m across the rupture trace of the 2010 M7.2 El Mayor-Cucapa earthquake to characterize the physical damage and chemical transformations associated with this active seismic source. In addition to quantifying fracture intensity from macroscopic analysis, we cut a continuous thin section through the fault core and from various samples in the damage zone, and ran each sample for XRD analyses for clay mineralogy, XRF for bulk geochemical analyses, and bulk and grain density from which porosity and volumetric strain were derived. The parent rock is a hydrothermally-altered biotite tonalite, with biotite partially altered to chlorite. The presence of epidote with chlorite suggests that these rocks were subjected to relatively high temperatures of 300-400° C. Adjacent to the outermost damage zone is a chaotic breccia zone with distinct chemical and physical characteristics, indicating possible connection to an ancestral fault to the southwest. The damage zone consists of an outer zone of protocataclasite, which grades inward towards mesocataclasite with seams of ultracataclasite. The fault core is anomalous in that it is largely composed of a sliver of marble that has been translated along the fault, so direct comparison with the damage zone is impaired. From collected data, we observe that chloritization increases into the breccia and damage zones, as does the presence of illite. Porosity reaches maximum values in the damage zone adjacent to the core, and closely follows trends in fracture intensity. Statistically significant gains in Mg, Na, K, Mn, and total bulk mass occurred within the inner damage zone, with losses of Ca and P mass, which led to the formation of chlorite and albite. The outer damage zone displays gains in Mg and Na mass with losses in Ca and P mass. The breccia zone shows gains in mass of Mg and Mn and loss in total bulk mass. A gain in LOI in both the

  9. Core damage frequency estimation using accident sequence precursor data: 1990-1993

    International Nuclear Information System (INIS)

    Martz, H.F.

    1998-01-01

    The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

  10. Precursors to potential severe core damage accidents: 1992, A status report

    International Nuclear Information System (INIS)

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N.; Dolan, B.W.; Jansen, J.M.; Minarick, J.W.; Lau, W.; Salyer, W.D.

    1993-12-01

    Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 x 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process

  11. 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

  12. Precursors to potential severe core damage accidents. A status report, 1982--1983

    Energy Technology Data Exchange (ETDEWEB)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W. [and others

    1997-04-01

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  13. Precursors to potential severe core damage accidents. A status report, 1982--1983

    International Nuclear Information System (INIS)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W.

    1997-04-01

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  14. Analysis of core damage frequency due to external events at the DOE [Department of Energy] N-Reactor

    International Nuclear Information System (INIS)

    Lambright, J.A.; Bohn, M.P.; Daniel, S.L.; Baxter, J.T.; Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P.; Brosseau, D.A.

    1990-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-11-01

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

  16. Status of the TMI-2 core: a review of damage assessments

    International Nuclear Information System (INIS)

    Croucher, D.W.

    1981-01-01

    Assessments of the damage within the core of the Three Mile Island Unit 2 reactor, performed by reconstructing the transient thermal-hydraulic sequence of events, estimating the amount of hydrogen generation, and evaluating the amount of fission products released, are reviewed and summarized. Minimum and maximum bounds of damage to the core are identified

  17. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    Bachrata, A.

    2012-01-01

    The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

  18. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  19. 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

  20. 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

  1. Field and laboratory investigations of coring-induced damage in core recovered from Marker Bed 139 at the waste isolation pilot plant underground facility

    International Nuclear Information System (INIS)

    Holcomb, D.J.; Zeuch, D.H.; Morin, K.; Hardy, R.; Tormey, T.V.

    1995-09-01

    A combined laboratory and field investigation was carried out to determine the extent of coring-induced damage done to samples cored from Marker Bed 139 at the WIPP site. Coring-induced damage, if present, has the potential to significantly change the properties of the material used for laboratory testing relative to the in situ material properties, resulting in misleading conclusions. In particular, connected, crack-like damage could make the permeability of cored samples orders of magnitude greater than the in situ permeabilities. Our approach compared in situ velocity and resistivity measurements with laboratory measurements of the same properties. Differences between in situ and laboratory results could be attributed to differences in the porosity due to cracks. The question of the origin of the changes could not be answered directly from the results of the measurements. Pre-existing cracks, held closed by the in situ stress, could open when the core was cut free, or new cracks could be generated by coring-induced damage. We used core from closely spaced boreholes at three orientations (0 degree, ±45 degrees relative to vertical) to address the origin of cracks. The absolute orientation of pre-existing cracks would be constant, independent of the borehole orientation. In contrast, cracks induced by coring were expected to show an orientation dependent on that of the source borehole

  2. Field and laboratory investigations of coring-induced damage in core recovered from Marker Bed 139 at the waste isolation pilot plant underground facility

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, D.J.; Zeuch, D.H.; Morin, K.; Hardy, R.; Tormey, T.V.

    1995-09-01

    A combined laboratory and field investigation was carried out to determine the extent of coring-induced damage done to samples cored from Marker Bed 139 at the WIPP site. Coring-induced damage, if present, has the potential to significantly change the properties of the material used for laboratory testing relative to the in situ material properties, resulting in misleading conclusions. In particular, connected, crack-like damage could make the permeability of cored samples orders of magnitude greater than the in situ permeabilities. Our approach compared in situ velocity and resistivity measurements with laboratory measurements of the same properties. Differences between in situ and laboratory results could be attributed to differences in the porosity due to cracks. The question of the origin of the changes could not be answered directly from the results of the measurements. Pre-existing cracks, held closed by the in situ stress, could open when the core was cut free, or new cracks could be generated by coring-induced damage. We used core from closely spaced boreholes at three orientations (0{degree}, {plus_minus}45{degrees} relative to vertical) to address the origin of cracks. The absolute orientation of pre-existing cracks would be constant, independent of the borehole orientation. In contrast, cracks induced by coring were expected to show an orientation dependent on that of the source borehole.

  3. 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

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

    International Nuclear Information System (INIS)

    Harper, F.T.

    1986-11-01

    This document contains the accident sequence analyses for Surry, Unit 1; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission (NRC). 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 Surry, Unit 1, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provide additional insights regarding the dominant contributors to the Surry core damage frequency estimate. The numerical results are driven to some degree by modeling assumptions and data selection for issues such as reactor coolant pump seal LOCAs, common cause failure probabilities, and plant response to station blackout and loss of electrical bust initiators. The sensitivity studies explore the impact of alternate theories and data on these issues

  5. Re criticality assessment following reactor core damage in Fukushima unit 2

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

    2012-01-01

    Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

  6. Precursors to potential severe core damage accidents: 1995 A status report

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.

    1997-04-01

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  7. Precursors to potential severe core damage accidents: 1995 A status report

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A. [and others

    1997-04-01

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  8. Management of radioactive waste from a major core damage in a BWR power plant

    International Nuclear Information System (INIS)

    Elkert, J.; Christensen, H.; Torstenfelt, B.

    1990-01-01

    Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

  9. 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

  10. Technical Note: Does Core Inflation Help Forecast Total Inflation? Evidence from Colombia

    OpenAIRE

    John Thornton

    1998-01-01

    In Colombia core and total inflation are both (1) series, and core inflation is cointegrated with total inflation. Granger causality tests using error correction methodology indicate that divergence of total inflation from core inflation is quickly revers

  11. Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

    International Nuclear Information System (INIS)

    Wider, H.

    2005-01-01

    New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)

  12. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  13. 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)

  14. Contribution of endogenous and exogenous damage to the total radiation-induced damage in the bacterial spore

    International Nuclear Information System (INIS)

    Jacobs, G.P.; Samuni, A.; Czapski, G.

    1980-01-01

    Radical scavengers such as polyethylene glycol 4000 and bovine albumin have been used to define the contribution of exogenous and endogenous damage to the total radiation-induced damage in aqueous buffered suspensions of Bacillus pumilus spores. The results indicate that this damage in the bacterial spore is predominantly endogenous

  15. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  16. Precursors to potential severe core damage accidents: 1996. A status report. Volume 25

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D.; Dolan, B.W.; Minarick, J.W.

    1997-12-01

    This report describes the 14 operational events in 1996 that affected 13 commercial light-water reactors and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1996 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1995 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  17. Precursors to potential severe core damage accidents: 1997 - A status report. Volume 26

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D.; Dolan, B.W.; Minarick, J.W.

    1998-11-01

    This report describes the five operational events in 1997 that affected five commercial light-water reactors (LWRs) and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1997 licensee event reports from commercial LWRs to identify those events that could be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1996 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  18. Radiation damage prediction system using damage function

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Mori, Seiji

    1979-01-01

    The irradiation damage analysis system using a damage function was investigated. This irradiation damage analysis system consists of the following three processes, the unfolding of a damage function, the calculation of the neutron flux spectrum of the object of damage analysis and the estimation of irradiation effect of the object of damage analysis. The damage function is calculated by applying the SAND-2 code. The ANISN and DOT3, 5 codes are used to calculate neutron flux. The neutron radiation and the allowable time of reactor operation can be estimated based on these calculations of the damage function and neutron flux. The flow diagram of the process of analyzing irradiation damage by a damage function and the flow diagram of SAND-2 code are presented, and the analytical code for estimating damage, which is determined with a damage function and a neutron spectrum, is explained. The application of the irradiation damage analysis system using a damage function was carried out to the core support structure of a fast breeder reactor for the damage estimation and the uncertainty evaluation. The fundamental analytical conditions and the analytical model for this work are presented, then the irradiation data for SUS304, the initial estimated values of a damage function, the error analysis for a damage function and the analytical results are explained concerning the computation of a damage function for 10% total elongation. Concerning the damage estimation of FBR core support structure, the standard and lower limiting values of damage, the permissible neutron flux and the allowable years of reactor operation are presented and were evaluated. (Nakai, Y.)

  19. Precursors to potential severe core damage accidents: 1992, A status report. Volume 17, Main report and Appendix A

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Jansen, J.M.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States); Lau, W.; Salyer, W.D. [Reliability and Performance Associates (United States)

    1993-12-01

    Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 {times} 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process.

  20. A backward method to estimate the Dai-ichi reactor core damage using radiation exposure in the environment

    International Nuclear Information System (INIS)

    PM Udiyani; S Kuntjoro; S Widodo

    2016-01-01

    The Fukushima accident resulted in the melting of the reactor core due to loss of supply of coolant when the reactor stopped from operating conditions. The earthquake and tsunami caused loss of electricity due to the flooding that occurred in the reactor. The absence of the coolant supply after reactor shutdown resulted in heat accumulation, causing the temperature of the fuel to rise beyond its melting point. In the early stages of the accident, operator could not determine the severity of the accident and the percentage of the reactor core damaged. The available data was based on the radiation exposure in the environment that was reported by the authorities. The aim of this paper is to determine the severity of the conditions in the reactor core based on the radiation doses measured in the environment. The method is performed by backward counting based on the measuring radiation exposure and radionuclides releases source term. The calculation was performed by using the PC-COSYMA code. The results showed that the core damage fraction at Dai-ichi Unit 1 was 70%, and the resulting individual effective dose in the exclusion area is 401 mSv, while the core damage fraction at Unit 2 was 30%, and the resulting individual effective dose was 9.1 mSv, while for Unit 3, the core damage fraction was 25% for an individual effective dose of 92.2 mSv. The differences between the results of the calculation for estimation of core damage proposed in this paper with the previously reported results is probably caused by the applied model for assessment, differences in postulations and assumptions, and the incompleteness of the input data. This difference could be reduced by performing calculations and simulations for more varied assumptions and postulations. (author)

  1. An estimation of core damage frequency of a pressurized water reactor during mid-loop operation

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    2004-01-01

    The core damage frequency during mid-loop operation of a Westinghouse designed 3-loop Pressurizer Water Reactor (PWR) due to loss of Residual Heat Removal (RHR) events was assessed. The assessment considers two types of outages (refueling and drained maintenance), and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events was identified and human error probabilities were quantified using HCR and THERP model. The result showed that the core damage frequency due to loss of RHR events during mid-loop operation is 3.1x10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering mid-loop operation. The establishment of reflux cooling, i.e. decay heat removal through steam generator secondary side also plays important role in mitigating the loss of RHR events. (author)

  2. Visualization of Heat Transfer and Core Damage With RGUI 1.5

    International Nuclear Information System (INIS)

    Mesina, George L.

    2002-01-01

    Graphical User Interfaces (GUI) have become an integral and essential part of computer software. In the ever-changing world of computing, they provide the user with a valuable means to learn, understand, and use the application software while also helping applications adapt to and span different computing paradigms, such as different operating systems. For these reasons, GUI development for nuclear plant analysis programs has been ongoing for a decade and a half and much progress has been made. With the development of codes such as RELAP5-3D [1] and SCDAP/RELAP5 that have multi-dimensional modeling capability, it has become necessary to represent three-dimensional, calculated data. The RELAP5-3D Graphical User Interface (RGUI) [4] was designed specifically for this purpose. It reduces the difficulty of analyzing complex three-dimensional models and enhances the analysts' ability to recognize plant behavior visually. Previous versions of RGUI [5] focused on visualizing reactor coolant behavior during a simulated transient or accident. Recent work has extended RGUI to display two other phenomena, heat transfer and core damage. Heat transfer is depicted through the visualization of RELAP5-3D heat structures. Core damage is visualized by displaying fuel rods and other core structures in a reactor vessel screen. Conditions within the core are displayed via numerical results and color maps. These new features of RGUI 1.5 are described and illustrated. (authors)

  3. Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

    International Nuclear Information System (INIS)

    Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

    1985-11-01

    A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

  4. Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vinnikov, B. [National Research Centre Kurchatov Inst., 1, Kurchatov Square, Moscow, 123 182 (Russian Federation); NRC Kurchatov Inst. (Russian Federation)

    2012-07-01

    According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

  5. Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

    International Nuclear Information System (INIS)

    Vinnikov, B.

    2012-01-01

    According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

  6. Core damage frequency perspectives based on IPE results

    International Nuclear Information System (INIS)

    Dingman, S.E.; Camp, A.L.; LaChance, J.L.; Drouin, M.T.

    1996-01-01

    In November 1988, the US Nuclear Regulatory Commission (NRC) issued Generic Letter 88-20 requesting that all licensees perform an individual Plant Examination (IPE) to identify any plant-specific vulnerability to severe accidents and report the results to the Commission. This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs

  7. Core damage frequency observations and insights of LWRs based on the IPEs

    Energy Technology Data Exchange (ETDEWEB)

    Dingman, S.E.; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Drouin, M.T. [and others

    1995-04-01

    Seventy-eight plants are expected to submit Individual Plant Examinations (IPEs) for severe accident vulnerabilities to the US Nuclear Regulatory Commission (NRC). The majority of the plants have elected to perform full Level 1 probabilistic risk assessments (PRAs) to meet the intent of the IPEs. Because of this, it is possible to compare the results from the IPE submittals to determine general observations and {open_quotes}lessons learned{close_quotes} from the IPEs. The IPE Insights Program is performing this evaluation, and preliminary results are presented in this paper. The core damage frequency and core damage sequences are identified and compared for pressurized water reactors and boiling water reactors. Examination of the results indicates that variations among plant results are due to a combination of actual plant design/operational features and analysis approaches. The findings are consistent with previous NRC studies, such as WASH-1400 and NUREG-1150.

  8. Core damage frequency observations and insights of LWRs based on the IPEs

    International Nuclear Information System (INIS)

    Dingman, S.E.; Camp, A.L.; Drouin, M.T.; Kolaczkowski, A.; Darby, J.; LaChance, J.L.; Yakle, J.

    1995-01-01

    Seventy-eight plants are expected to submit Individual Plant Examinations (IPEs) for severe accident vulnerabilities to the U.S. Nuclear Regulatory Commission (NRC). The majority of the plants have elected to perform full Level 1 probabilistic risk assessments (PRAs) to meet the intent of the IPES. Because of this, it is possible to compare the results from the IPE submittals to determine general observations and open-quotes lessons learnedclose quotes from the IPES. The IPE Insights Program is performing this evaluation, and preliminary results are presented in this paper. The core damage frequency and core damage sequences are identified and compared for pressurized water reactors and boiling water reactors. Examination of the results indicates that variations among plant results are due to a combination of actual plant design/operational features and analysis approaches. The findings are consistent with previous NRC studies, such as WASH-1400 and NUREG-1 150

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

    International Nuclear Information System (INIS)

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

    2009-03-01

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

  10. Preparations to load, transport, receive, and store the damaged TMI-2 [Three Mile Island] reactor core

    International Nuclear Information System (INIS)

    Reno, H.W.; Schmitt, R.C.; Quinn, G.J.; Ayers, A.L. Jr.; Lilburn, B.J. Jr.; Uhl, D.L.

    1986-03-01

    The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations for transporting the core debris from TMI to INEL and receiving and storing that material at INEL. Issues discussed include interfacing of equipment and facilities at TMI, loading operations, transportation activities using a newly designed cask, receiving and storing operations at INEL, and criticality control during storage. Key to the transportation effort was designing, testing, fabricating, and licensing two rail casks which individually provide double containment of the damaged fuel. 27 figs

  11. Drilling induced damage of core samples. Evidences from laboratory testing and numerical modelling

    International Nuclear Information System (INIS)

    Lanaro, Flavio

    2008-01-01

    Extensive sample testing in uniaxial and Brazilian test conditions were carried out for the Shobasama and MIU Research Laboratory Site (Gifu Pref., Japan). The compressive and tensile strength of the samples was observed to be negatively correlated to the in-situ stress components. Such correlation was interpreted as stress-release induced sample damage. Similar stress conditions were then numerically simulated by means of the BEM-DDM code FRACOD 2D in plane strain conditions. This method allows for explicitly consider the influence of newly initiated or propagating fractures on the stress field and deformation of the core during drilling process. The models show that already at moderate stress levels some fracturing of the core during drilling might occur leading to reduced laboratory strength of the samples. Sample damage maps were produced independently from the laboratory test results and from the numerical models and show good agreement with each other. (author)

  12. Sensitivity Analysis of Core Damage by Loss of Auxiliary Feed Water during the Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Woo Jae; Chung, Soon Il; Hwang, Su Hyun; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech., Yongin (Korea, Republic of); Yun, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the reactor core damage time for OPR1000 type Nuclear Power Plant (NPP) was analyzed to develop a strategy to handle ELAP and to apply to the EOP. The reactor core damage time in the ELAP condition was calculated according to the time of Auxiliary Feedwater (AFW) loss. Fukushima accident was caused by long hours of Station Black Out (SBO) caused by natural disaster beyond Design Based Accident (DBA) criteria. It led to the reactor core damage. After the accident, the regulatory authorities of each country (Japan, US, EU, IAEA, and etc.) recommended developing the necessary systems and strategies in order to cover up the Extended Loss of All AC Power (ELAP) such as one occurred in the Fukushima accident. And the need of procedure or guideline to cope with ELAP has been raised through the stress test for Wolsong Nuclear Power Plant unit 1. Current Emergency Operating Procedures (EOP) used in domestic nuclear power plant are seemed to be insufficient to cope with ELAP. Therefore, it has been required to be improved. As the result, the time of AFW loss in the ELAP condition influences greatly on core damage time.

  13. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N.; Dolan, B.W.; Minarick, J.W.

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1

  14. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

  15. Core damage frequency (reactor design) perspectives based on IPE results

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-01-01

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed

  16. Core food of the French food supply: second Total Diet Study.

    Science.gov (United States)

    Sirot, V; Volatier, J L; Calamassi-Tran, G; Dubuisson, C; Menard, C; Dufour, A; Leblanc, J C

    2009-05-01

    As first described in the 1980s, the core food intake model allows a precise assessment of dietary nutrient intake and dietary exposure to contaminants insofar as it reflects the eating habits of a target population and covers the most important foods in terms of consumption, selected nutrient and contaminant contribution. This model has been used to set up the sampling strategy of the second French Total Diet Study (TDS) with the aim of obtaining a realistic panorama of nutrient intakes and contaminant exposure for the whole population, useful for quantitative risk assessment. Data on consumption trends and eating habits from the second French individual food consumption survey (INCA2) as well as data from a 2004 purchase panel of French households (SECODIP) were used to identify the core foods to be sampled. A total of 116 core foods on a national scale and 70 core foods on a regional scale were selected according to (1) the consumption data for adults and children, (2) their consumer rates, and (3) their high contribution to exposure to one or more contaminants of interest. Foods were collected in eight French regions (36 cities) and prepared 'as consumed' to be analysed for their nutritional composition and contamination levels. A total of 20 280 different food products were purchased to make up the 1352 composite samples of core foods to be analysed for additives, environmental contaminants, pesticide residues, trace elements and minerals, mycotoxins and acrylamide. The establishment of such a sampling plan is essential for effective, high-quality monitoring of dietary exposure from a public health point of view.

  17. Pulsed total dose damage effect experimental study on EPROM

    International Nuclear Information System (INIS)

    Luo Yinhong; Yao Zhibin; Zhang Fengqi; Guo Hongxia; Zhang Keying; Wang Yuanming; He Baoping

    2011-01-01

    Nowadays, memory radiation effect study mainly focus on functionality measurement. Measurable parameters is few in china. According to the present situation, threshold voltage testing method was presented on floating gate EPROM memory. Experimental study of pulsed total dose effect on EPROM threshold voltage was carried out. Damage mechanism was analysed The experiment results showed that memory cell threshold voltage negative shift was caused by pulsed total dose, memory cell threshold voltage shift is basically coincident under steady bias supply and no bias supply. (authors)

  18. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effect of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  19. A core hSSB1–INTS complex participates in the DNA damage response

    OpenAIRE

    Zhang, Feng; Ma, Teng; Yu, Xiaochun

    2013-01-01

    Human single-stranded DNA-binding protein 1 (hSSB1) plays an important role in the DNA damage response and the maintenance of genomic stability. It has been shown that the core hSSB1 complex contains hSSB1, INTS3 and C9orf80. Using protein affinity purification, we have identified integrator complex subunit 6 (INTS6) as a major subunit of the core hSSB1 complex. INTS6 forms a stable complex with INTS3 and hSSB1 both in vitro and in vivo. In this complex, INTS6 directly interacts with INTS3. I...

  20. Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

    International Nuclear Information System (INIS)

    Chang, L.K.; Koenig, J.F.; Porter, D.L.

    1987-01-01

    In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

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

    International Nuclear Information System (INIS)

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

    1989-08-01

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

  2. Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

    International Nuclear Information System (INIS)

    Kim, See Darl; Kim, Dong Ha

    1998-04-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

  3. FANCI Regulates Recruitment of the FA Core Complex at Sites of DNA Damage Independently of FANCD2.

    Directory of Open Access Journals (Sweden)

    Maria Castella

    2015-10-01

    Full Text Available The Fanconi anemia (FA-BRCA pathway mediates repair of DNA interstrand crosslinks. The FA core complex, a multi-subunit ubiquitin ligase, participates in the detection of DNA lesions and monoubiquitinates two downstream FA proteins, FANCD2 and FANCI (or the ID complex. However, the regulation of the FA core complex itself is poorly understood. Here we show that the FA core complex proteins are recruited to sites of DNA damage and form nuclear foci in S and G2 phases of the cell cycle. ATR kinase activity, an intact FA core complex and FANCM-FAAP24 were crucial for this recruitment. Surprisingly, FANCI, but not its partner FANCD2, was needed for efficient FA core complex foci formation. Monoubiquitination or ATR-dependent phosphorylation of FANCI were not required for the FA core complex recruitment, but FANCI deubiquitination by USP1 was. Additionally, BRCA1 was required for efficient FA core complex foci formation. These findings indicate that FANCI functions upstream of FA core complex recruitment independently of FANCD2, and alter the current view of the FA-BRCA pathway.

  4. Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

    Energy Technology Data Exchange (ETDEWEB)

    Fok, Alex

    2013-10-30

    The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

  5. Precursors to potential severe core damage accidents: 1994, a status report. Volume 21: Main report and appendices A--H

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N.; Dolan, B.W.; Minarick, J.W.

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1

  6. An estimation of core damage frequency of a pressurized water reactor during midloop operation due to loss of residual heat removal

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    1995-01-01

    The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 x 10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation

  7. IAEA Regional Workshop on Development and Validation of EOP/AMG for Effective Prevention/Mitigation of Severe Core Damage

    International Nuclear Information System (INIS)

    1999-01-01

    Materials of the IAEA Regional Workshop contain 24 presented lectures. Authors deal with development and validation of emergency operating procedures as well as with accident management guidelines (EOP/AMG) for effective prevention and mitigation of severe core damage

  8. Geometry of the Nojima fault at Nojima-Hirabayashi, Japan - I. A simple damage structure inferred from borehole core permeability

    Science.gov (United States)

    Lockner, David A.; Tanaka, Hidemi; Ito, Hisao; Ikeda, Ryuji; Omura, Kentaro; Naka, Hisanobu

    2009-01-01

    The 1995 Kobe (Hyogo-ken Nanbu) earthquake, M = 7.2, ruptured the Nojima fault in southwest Japan. We have studied core samples taken from two scientific drillholes that crossed the fault zone SW of the epicentral region on Awaji Island. The shallower hole, drilled by the Geological Survey of Japan (GSJ), was started 75 m to the SE of the surface trace of the Nojima fault and crossed the fault at a depth of 624 m. A deeper hole, drilled by the National Research Institute for Earth Science and Disaster Prevention (NIED) was started 302 m to the SE of the fault and crossed fault strands below a depth of 1140 m. We have measured strength and matrix permeability of core samples taken from these two drillholes. We find a strong correlation between permeability and proximity to the fault zone shear axes. The half-width of the high permeability zone (approximately 15 to 25 m) is in good agreement with the fault zone width inferred from trapped seismic wave analysis and other evidence. The fault zone core or shear axis contains clays with permeabilities of approximately 0.1 to 1 microdarcy at 50 MPa effective confining pressure (10 to 30 microdarcy at in situ pressures). Within a few meters of the fault zone core, the rock is highly fractured but has sustained little net shear. Matrix permeability of this zone is approximately 30 to 60 microdarcy at 50 MPa effective confining pressure (300 to 1000 microdarcy at in situ pressures). Outside this damage zone, matrix permeability drops below 0.01 microdarcy. The clay-rich core material has the lowest strength with a coefficient of friction of approximately 0.55. Shear strength increases with distance from the shear axis. These permeability and strength observations reveal a simple fault zone structure with a relatively weak fine-grained core surrounded by a damage zone of fractured rock. In this case, the damage zone will act as a high-permeability conduit for vertical and horizontal flow in the plane of the

  9. Estimation of core-damage frequency to evolutionary ALWR [advanced light water reactor] due to seismic initiating events: Task 4.3.3

    International Nuclear Information System (INIS)

    Brooks, R.D.; Harrison, D.G.; Summitt, R.L.

    1990-04-01

    The Electric Power Research Institute (EPRI) is presently developing a requirements document for the design of advanced light water reactors (ALWRs). One of the basic goals of the EPRI ALWR Requirements Document is that the core-damage frequency for an ALWR shall be less than 1.0E-5. To aid in this effort, the Department of Energy's Advanced Reactor Severe Accident Program (ARSAP) initiated a functional probabilistic risk assessment (PRA) to determine how effectively the evolutionary plant requirements contained in the existing EPRI Requirements Document assure that this safety goal will be met. This report develops an approximation of the core-damage frequency due to seismic events for both evolutionary plant designs (pressurized-water reactor (PWR) and boiling-water reactor(BWR)) as modeled in the corresponding functional PRAs. Component fragility values were taken directly form information which has been submitted for inclusion in Appendix A to Volume 1 of the EPRI Requirements Document. The results show a seismic core-damage frequency of 5.2E-6 for PWRS and 5.0E-6 for BWRs. Combined with the internal initiators from the functional PRAs, the overall core-damage frequencies are 6.0E-6 for the pwr and BWR, both of which satisfy the 1.0E-5 EPRI goal. In addition, site-specific considerations, such as more rigid components and less conservative fragility data and seismic hazard curves, may further reduce these frequencies. The effect of seismic events on structures are not addressed in this generic evaluation and should be addressed separately on a design-specific basis. 7 refs., 6 figs., 3 tabs

  10. Phenomena occurring in the reactor coolant system during severe core damage accidents

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1989-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

  11. Compendium of Total Ionizing Dose and Displacement Damage for Candidate Spacecraft Electronics for NASA

    Science.gov (United States)

    Cochran, Donna J.; Boutte, Alvin J.; Chen, Dakai; Pellish, Jonathan A.; Ladbury, Raymond L.; Casey, Megan C.; Campola, Michael J.; Wilcox, Edward P.; Obryan, Martha V.; LaBel, Kenneth A.; hide

    2012-01-01

    Vulnerability of a variety of candidate spacecraft electronics to total ionizing dose and displacement damage is studied. Devices tested include optoelectronics, digital, analog, linear, and hybrid devices.

  12. Survival of extensively damaged endodontically treated incisors restored with different types of posts-and-core foundation restoration material.

    Science.gov (United States)

    Lazari, Priscilla Cardoso; de Carvalho, Marco Aurélio; Del Bel Cury, Altair A; Magne, Pascal

    2018-05-01

    Which post-and-core combination will best improve the performance of extensively damaged endodontically treated incisors without a ferrule is still unclear. The purpose of this in vitro study was to investigate the restoration of extensively damaged endodontically treated incisors without a ferrule using glass-ceramic crowns bonded to various composite resin foundation restorations and 2 types of posts. Sixty decoronated endodontically treated bovine incisors without a ferrule were divided into 4 groups and restored with various post-and-core foundation restorations. NfPfB=no-ferrule (Nf) with glass-fiber post (Pf) and bulk-fill resin foundation restoration (B); NfPfP=no-ferrule (Nf) with glass-fiber post (Pf) and dual-polymerized composite resin core foundation restoration (P); NfPt=no-ferrule (Nf) with titanium post (Pt) and resin core foundation restoration; and NfPtB=no-ferrule (Nf) with titanium post (Pt) and bulk-fill resin core foundation restoration (B). Two additional groups from previously published data from the same authors (FPf=2mm of ferrule (F) and glass-fiber post (Pf) and composite resin core foundation restoration; and NfPf=no-ferrule (Nf) with glass-fiber post (Pf) and composite resin core foundation restoration), which were tested concomitantly and using the same experimental arrangement, were included for comparison. All teeth were prepared to receive bonded glass-ceramic crowns luted with dual-polymerized resin cement and were subjected to accelerated fatigue testing under submerged conditions at room temperature. Cyclic isometric loading was applied to the incisal edge at an angle of 30 degrees with a frequency of 5 Hz, beginning with a load of 100 N (5000 cycles). A 100-N load increase was applied every 15000 cycles. The specimens were loaded until failure or to a maximum of 1000 N (140000 cycles). The 6 groups (4 groups from the present study and 2 groups from the previously published study) were compared using the Kaplan-Meier survival

  13. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

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

    1995-07-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

  14. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal floods during mid-loop operations. Volume 4

    International Nuclear Information System (INIS)

    Kohut, P.

    1994-07-01

    The major objective of the Surry internal flood analysis was to provide an improved understanding of the core damage scenarios arising from internal flood-related events. The mean core damage frequency of the Surry plant due to internal flood events during mid-loop operations is 4.8E-06 per year, and the 5th and 95th percentiles are 2.2E-07 and 1.8E-05 per year, respectively. Some limited sensitivity calculations were performed on three plant improvement options. The most significant result involves modifications of intake-level structure on the canal, which reduced core damage frequency contribution from floods in mid-loop by about 75%

  16. Core damage vulnerability due to the loss of ESW [essential service water] systems at multiplant sites: An assessment and options

    International Nuclear Information System (INIS)

    Kohut, P.; Musicki, Z.; Fitzpatrick, R.

    1989-01-01

    The main objective of this study is to establish the core damage vulnerability caused by the failure of the ESW systems in multiplant units that have only two sw pumps per unit with crosstie capability. Design and operating data have been surveyed to derive system failure frequency. A core damage model is constructed including operating configurations, specific recovery actions, and time and leak rate dependent RCP seal LOCA model. The estimated CDF SW = 2.55 x 10 -4 /yr is significant indicating the potential vulnerability of this particular SW design arrangement. A number of different potential improvements have been considered. The addition of a swing pump serving both units is shown to have the most significant CDF reduction potential (∼50%) combined with advantageous cost/benefit aspects. 2 refs., 2 tabs

  17. 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)

  18. Recreational stimulants, herbal, and spice cannabis: The core psychobiological processes that underlie their damaging effects.

    Science.gov (United States)

    Parrott, Andrew C; Hayley, Amie C; Downey, Luke A

    2017-05-01

    Recreational drugs are taken for their positive mood effects, yet their regular usage damages well-being. The psychobiological mechanisms underlying these damaging effects will be debated. The empirical literature on recreational cannabinoids and stimulant drugs is reviewed. A theoretical explanation for how they cause similar types of damage is outlined. All psychoactive drugs cause moods and psychological states to fluctuate. The acute mood gains underlie their recreational usage, while the mood deficits on withdrawal explain their addictiveness. Cyclical mood changes are found with every central nervous system stimulant and also occur with cannabis. These mood state changes provide a surface index for more profound psychobiological fluctuations. Homeostatic balance is altered, with repetitive disturbances of the hypothalamic-pituitary-adrenal axis, and disrupted cortisol-neurohormonal secretions. Hence, these drugs cause increased stress, disturbed sleep, neurocognitive impairments, altered brain activity, and psychiatric vulnerability. Equivalent deficits occur with novel psychoactive stimulants such as mephedrone and artificial "spice" cannabinoids. These psychobiological fluctuations underlie drug dependency and make cessation difficult. Psychobiological stability and homeostatic balance are optimally restored by quitting psychoactive drugs. Recreational stimulants such as cocaine or MDMA (3.4-methylenedioxymethamphetamine) and sedative drugs such as cannabis damage human homeostasis and well-being through similar core psychobiological mechanisms. Copyright © 2017 John Wiley & Sons, Ltd.

  19. Recent Total Ionizing Dose and Displacement Damage Compendium of Candidate Electronics for NASA Space Systems

    Science.gov (United States)

    Cochran, Donna J.; Boutte, Alvin J.; Campola, Michael J.; Carts, Martin A.; Casey, Megan C.; Chen, Dakai; LaBel, Kenneth A.; Ladbury, Raymond L.; Lauenstein, Jean-Marie; Marshall, Cheryl J.; hide

    2011-01-01

    Vulnerability of a variety of candidate spacecraft electronics to total ionizing dose and displacement damage is studied. Devices tested include optoelectronics, digital, analog, linear bipolar devices, and hybrid devices.

  20. Metabolite Damage and Metabolite Damage Control in Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, Andrew D. [Horticultural Sciences Department and; Henry, Christopher S. [Mathematics and Computer Science Division, Argonne National Laboratory, Argonne, Illinois 60439, email:; Computation Institute, University of Chicago, Chicago, Illinois 60637; Fiehn, Oliver [Genome Center, University of California, Davis, California 95616, email:; de Crécy-Lagard, Valérie [Microbiology and Cell Science Department, University of Florida, Gainesville, Florida 32611, email: ,

    2016-04-29

    It is increasingly clear that (a) many metabolites undergo spontaneous or enzyme-catalyzed side reactions in vivo, (b) the damaged metabolites formed by these reactions can be harmful, and (c) organisms have biochemical systems that limit the buildup of damaged metabolites. These damage-control systems either return a damaged molecule to its pristine state (metabolite repair) or convert harmful molecules to harmless ones (damage preemption). Because all organisms share a core set of metabolites that suffer the same chemical and enzymatic damage reactions, certain damage-control systems are widely conserved across the kingdoms of life. Relatively few damage reactions and damage-control systems are well known. Uncovering new damage reactions and identifying the corresponding damaged metabolites, damage-control genes, and enzymes demands a coordinated mix of chemistry, metabolomics, cheminformatics, biochemistry, and comparative genomics. This review illustrates the above points using examples from plants, which are at least as prone to metabolite damage as other organisms.

  1. Managing water addition to a degraded core

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Assessment of damage potential to the TMI-2 lower head due to thermal attack by core debris

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Behling, S.R.; Broughton, J.M.

    1986-06-01

    Camera inspection of the Three Mile Island Unit 2 (TMI-2) inlet plenum region has shown that approximately 10 to 20 percent of the core material loading may have relocated to the lower plenum. Although vessel integrity was maintained, a question of primary concern is ''how close to vessel failure'' did this accident come. This report summarizes the results of thermal analyses aimed at assessing damage potential to the TMI-2 lower head and attached instrument penetration tubes due to thermal attack by hot core debris. Results indicate that the instrument penetration nozzles could have experienced melt failure at localized hot spot regions, with attendant debris drainage and plugging of the instrument lead tubes. However, only minor direct thermal attack of the vessel liner is predicted

  3. Compendium of Current Total Ionizing Dose and Displacement Damage Results from NASA Goddard Space Flight Center and NASA Electronic Parts and Packaging Program

    Science.gov (United States)

    Topper, Alyson D.; Campola, Michael J.; Chen, Dakai; Casey, Megan C.; Yau, Ka-Yen; Cochran, Donna J.; Label, Kenneth A.; Ladbury, Raymond L.; Mondy, Timothy K.; O'Bryan, Martha V.; hide

    2017-01-01

    Total ionizing dose and displacement damage testing was performed to characterize and determine the suitability of candidate electronics for NASA space utilization. Devices tested include optoelectronics, digital, analog, linear bipolar devices, and hybrid devices. Displacement Damage, Optoelectronics, Proton Damage, Single Event Effects, and Total Ionizing Dose.

  4. Compendium of Current Total Ionizing Dose and Displacement Damage Results from NASA GSFC and NEPP

    Science.gov (United States)

    Topper, Alyson D.; Campola, Michael J.; Chen, Dakai; Casey, Megan C.; Yau, Ka-Yen; Label, Kenneth A.; Cochran, Donna J.; O'Bryan, Martha V.

    2017-01-01

    Total ionizing dose and displacement damage testing was performed to characterize and determine the suitability of candidate electronics for NASA space utilization. Devices tested include opto-electronics, digital, analog, linear bipolar devices, and hybrid devices.

  5. Concept and methodology for evaluating core damage frequency considering failure correlation at multi units and sites and its application

    Energy Technology Data Exchange (ETDEWEB)

    Ebisawa, K.; Teragaki, T.; Nomura, S. [Former Incorporated Administrative Agency, Japan Nuclear Safety Organization (Japan); Abe, H., E-mail: Hiroshi_abe@nsr.go.jp [Former Incorporated Administrative Agency, Japan Nuclear Safety Organization (Japan); Shigemori, M.; Shimomoto, M. [Mizuho Information & Research Institute, 2-3, Kanda-Nishikicho, Chiyoda-ku, Tokyo (Japan)

    2015-07-15

    Highlights: • We develop a method to evaluate CDF considering failure correlation at multi units. • We develop a procedure to evaluate correlation coefficient between multi components. • We evaluate CDF at two different BWR units using correlation coefficients. • We confirm the validity of method and correlation coefficient through the evaluation. - Abstract: The Tohoku earthquake (Mw9.0) occurred on March 11, 2011 and caused a large tsunami. The Fukushima Daiichi Nuclear Power Plant with six units were overwhelmed by the tsunami and core damage occurred. Authors proposed the concept and method for evaluating core damage frequency (CDF) considering failure correlation at the multi units and sites. Based on the above method, one of authors developed the procedure for evaluating the failure correlation coefficient and response correlation coefficient between the multi components under the strong seismic motion. These method and failure correlation coefficients were applied to two different BWR units and their CDF was evaluated by seismic probabilistic risk assessment technology. Through this quantitative evaluation, the validity of the method and failure correlation coefficient was confirmed.

  6. Concept and methodology for evaluating core damage frequency considering failure correlation at multi units and sites and its application

    International Nuclear Information System (INIS)

    Ebisawa, K.; Teragaki, T.; Nomura, S.; Abe, H.; Shigemori, M.; Shimomoto, M.

    2015-01-01

    Highlights: • We develop a method to evaluate CDF considering failure correlation at multi units. • We develop a procedure to evaluate correlation coefficient between multi components. • We evaluate CDF at two different BWR units using correlation coefficients. • We confirm the validity of method and correlation coefficient through the evaluation. - Abstract: The Tohoku earthquake (Mw9.0) occurred on March 11, 2011 and caused a large tsunami. The Fukushima Daiichi Nuclear Power Plant with six units were overwhelmed by the tsunami and core damage occurred. Authors proposed the concept and method for evaluating core damage frequency (CDF) considering failure correlation at the multi units and sites. Based on the above method, one of authors developed the procedure for evaluating the failure correlation coefficient and response correlation coefficient between the multi components under the strong seismic motion. These method and failure correlation coefficients were applied to two different BWR units and their CDF was evaluated by seismic probabilistic risk assessment technology. Through this quantitative evaluation, the validity of the method and failure correlation coefficient was confirmed

  7. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  8. Multi-scale fracture damage associated with underground chemical explosions

    Science.gov (United States)

    Swanson, E. M.; Sussman, A. J.; Wilson, J. E.; Townsend, M. J.; Prothro, L. B.; Gang, H. E.

    2018-05-01

    Understanding rock damage induced by explosions is critical for a number of applications including the monitoring and verification of underground nuclear explosions, mine safety issues, and modeling fluid flow through fractured rock. We use core observations, televiewer logs, and thin section observations to investigate fracture damage associated with two successive underground chemical explosions (SPE2 and SPE3) in granitic rock at both the mesoscale and microscale. We compare the frequency and orientations of core-scale fractures, and the frequency of microfractures, between a pre-experiment core and three post-experiment cores. Natural fault zones and explosion-induced fractures in the vicinity of the explosive source are readily apparent in recovered core and in thin sections. Damage from faults and explosions is not always apparent in fracture frequency plots from televiewer logs, although orientation data from these logs suggests explosion-induced fracturing may not align with the pre-existing fracture sets. Core-scale observations indicate the extent of explosion-induced damage is 10.0 m after SPE2 and 6.8 m after SPE3, despite both a similar size and location for both explosions. At the microscale, damage is observed to a range distance of 10.2 ± 0.9 m after SPE2, and 16.6 ± 0.9 and 11.2 ± 0.6 in two different cores collected after SPE3. Additional explosion-induced damage, interpreted to be the result of spalling, is readily apparent near the surface, but only in the microfracture data. This depth extent and intensity of damage in the near-surface region also increased after an additional explosion. This study highlights the importance of evaluating structural damage at multiple scales for a more complete characterization of the damage, and particularly shows the importance of microscale observations for identifying spallation-induced damage.

  9. [Operative treatment strategies for multiple trauma patients : early total care versus damage control].

    Science.gov (United States)

    Klüter, T; Lippross, S; Oestern, S; Weuster, M; Seekamp, A

    2013-09-01

    The treatment of multiple trauma patients is a great challenge for an interdisciplinary team. After preclinical care and subsequent treatment in the emergency room the order of the interventions is prioritized depending of the individual risk stratification. For planning the surgery management it is essential to distinguish between absolutely essential operations to prevent life-threatening situations for the patient and interventions with shiftable indications, depending on the general condition of the patient. All interventions need to be done without causing significant secondary damage to prohibit hyperinflammation and systemic inflammatory response syndrome. The challenge consists in determination of the appropriate treatment at the right point in time. In general the early primary intervention, early total care, is differentiated from the damage control concept.

  10. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  11. The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

    International Nuclear Information System (INIS)

    Martz, H.F.

    1995-05-01

    It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

  12. Optimized core loading sequence for Ukraine WWER-1000 reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

  13. The IPE Database: providing information on plant design, core damage frequency and containment performance

    International Nuclear Information System (INIS)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T.; Su, T.; Danziger, L.

    1996-01-01

    A database, called the IPE Database has been developed that stores data obtained from the Individual Plant Examinations (IPEs) which licensees of nuclear power plants have conducted in response to the Nuclear Regulatory Commission's (NRC) Generic Letter GL88-20. The IPE Database is a collection of linked files which store information about plant design, core damage frequency (CDF), and containment performance in a uniform, structured way. The information contained in the various files is based on data contained in the IPE submittals. The information extracted from the submittals and entered into the IPE Database can be manipulated so that queries regarding individual or groups of plants can be answered using the IPE Database

  14. Protection of radiation induced DNA and membrane damages by total triterpenes isolated from Ganoderma lucidum (Fr.) P. Karst.

    Science.gov (United States)

    Smina, T P; Maurya, D K; Devasagayam, T P A; Janardhanan, K K

    2015-05-25

    The total triterpenes isolated from the fruiting bodies of Ganoderma lucidum was examined for its potential to prevent γ-radiation induced membrane damage in rat liver mitochondria and microsomes. The effects of total triterpenes on γ-radiation-induced DNA strand breaks in pBR 322 plasmid DNA in vitro and human peripheral blood lymphocytes ex vivo were evaluated. The protective effect of total triterpenes against γ-radiation-induced micronuclei formations in mice bone marrow cells in vivo were also evaluated. The results indicated the significant effectiveness of Ganoderma triterpenes in protecting the DNA and membrane damages consequent to the hazardous effects of radiation. The findings suggest the potential use of Ganoderma triterpenes in radio therapy. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  15. Recent Total Ionizing Dose Results and Displacement Damage Results for Candidate Spacecraft Electronics for NASA

    Science.gov (United States)

    Cochran, Donna J.; Buchner, Stephen P.; Irwin, Tim L.; LaBel, Kenneth A.; Marshall, Cheryl J.; Reed, Robert A.; Sanders, Anthony B.; Hawkins, Donald K.; Flanigan, Ryan J.; Cox, Stephen R.

    2005-01-01

    We present data on the vulnerability of a variety of candidate spacecraft electronics to total ionizing dose and displacement damage. Devices tested include optoelectronics, digital, analog, linear bipolar devices, hybrid devices, Analog-to- Digital Converters (ADCs), and Digital-to-Analog Converters (DACs), among others. T

  16. Enchanced total dose damage in junction field effect transistors and related linear integrated circuits

    International Nuclear Information System (INIS)

    Flament, O.; Autran, J.L.; Roche, P.; Leray, J.L.; Musseau, O.

    1996-01-01

    Enhanced total dose damage of Junction Field-effect Transistors (JFETs) due to low dose rate and/or elevated temperature has been investigated for elementary p-channel structures fabricated on bulk and SOI substrates as well as for related linear integrated circuits. All these devices were fabricated with conventional junction isolation (field oxide). Large increases in damage have been revealed by performing high temperature and/or low dose rate irradiations. These results are consistent with previous studies concerning bipolar field oxides under low-field conditions. They suggest that the transport of radiation-induced holes through the oxide is the underlying mechanism. Such an enhanced degradation must be taken into account for low dose rate effects on linear integrated circuits

  17. Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

    1995-01-01

    This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

  18. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  19. Tailoring Sandwich Face/Core Interfaces for Improved Damage Tolerance

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Carlsson, Leif A.

    2010-01-01

    A face/core debond in a sandwich structure may propagate in the interface or kink into either the face or core. It is found that certain modifications of the face/core interface region influence the kinking behavior, which is studied experimentally in the present paper. A sandwich double cantilever....... The transition points where the crack kinks are identified and the influence of four various interface design modifications on the propagation path and fracture resistance are investigated....

  20. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-01-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation

  1. Review of the Oconee-3 probabilistic risk assessment: external events, core damage frequency. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, N.A.; Ilberg, D.; Xue, D.; Youngblood, R.; Reed, J.W.; McCann, M.; Talwani, T.; Wreathall, J.; Kurth, P.D.; Bandyopadhyay, K.

    1986-03-01

    A review of the Oconee-3 Probabilistic Risk Assessment (OPRA) was conducted with the broad objective of evaluating qualitatively and quantitatively (as much as possible) the OPRA assessment of the important sequences that are ''externally'' generated and lead to core damage. The review included a technical assessment of the assumptions and methods used in the OPRA within its stated objective and with the limited information available. Within this scope, BNL performed a detailed reevaluation of the accident sequences generated by internal floods and earthquakes and a less detailed review (in some cases a scoping review) for the accident sequences generated by fires, tornadoes, external floods, and aircraft impact. 12 refs., 24 figs., 31 tabs.

  2. Review of the Oconee-3 probabilistic risk assessment: external events, core damage frequency. Volume 2

    International Nuclear Information System (INIS)

    Hanan, N.A.; Ilberg, D.; Xue, D.

    1986-03-01

    A review of the Oconee-3 Probabilistic Risk Assessment (OPRA) was conducted with the broad objective of evaluating qualitatively and quantitatively (as much as possible) the OPRA assessment of the important sequences that are ''externally'' generated and lead to core damage. The review included a technical assessment of the assumptions and methods used in the OPRA within its stated objective and with the limited information available. Within this scope, BNL performed a detailed reevaluation of the accident sequences generated by internal floods and earthquakes and a less detailed review (in some cases a scoping review) for the accident sequences generated by fires, tornadoes, external floods, and aircraft impact. 12 refs., 24 figs., 31 tabs

  3. Characterization of total ionizing dose damage in COTS pinned photodiode CMOS image sensors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zujun, E-mail: wangzujun@nint.ac.cn; Ma, Wuying; Huang, Shaoyan; Yao, Zhibin; Liu, Minbo; He, Baoping; Sheng, Jiangkun; Xue, Yuan [State Key Laboratory of Intense Pulsed Radiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O.Box 69-10, Xi’an, Shaanxi 710024 (China); Liu, Jing [School of Materials Science and Engineering, Xiangtan University, Hunan (China)

    2016-03-15

    The characterization of total ionizing dose (TID) damage in COTS pinned photodiode (PPD) CMOS image sensors (CISs) is investigated. The radiation experiments are carried out at a {sup 60}Co γ-ray source. The CISs are produced by 0.18-μm CMOS technology and the pixel architecture is 8T global shutter pixel with correlated double sampling (CDS) based on a 4T PPD front end. The parameters of CISs such as temporal domain, spatial domain, and spectral domain are measured at the CIS test system as the EMVA 1288 standard before and after irradiation. The dark current, random noise, dark signal non-uniformity (DSNU), photo response non-uniformity (PRNU), overall system gain, saturation output, dynamic range (DR), signal to noise ratio (SNR), quantum efficiency (QE), and responsivity versus the TID are reported. The behaviors of the tested CISs show remarkable degradations after radiation. The degradation mechanisms of CISs induced by TID damage are also analyzed.

  4. NDE of Damage in Aircraft Flight Control Surfaces

    International Nuclear Information System (INIS)

    Hsu, David K.; Barnard, Daniel J.; Dayal, Vinay

    2007-01-01

    Flight control surfaces on an aircraft, such as ailerons, flaps, spoilers and rudders, are typically adhesively bonded composite or aluminum honeycomb sandwich structures. These components can suffer from damage caused by hail stone, runway debris, or dropped tools during maintenance. On composites, low velocity impact damages can escape visual inspection, whereas on aluminum honeycomb sandwich, budding failure of the honeycomb core may or may not be accompanied by a disbond. This paper reports a study of the damage morphology in such structures and the NDE methods for detecting and characterizing them. Impact damages or overload failures in composite sandwiches with Nomex or fiberglass core tend to be a fracture or crinkle or the honeycomb cell wall located a distance below the facesheet-to-core bondline. The damage in aluminum honeycomb is usually a buckling failure, propagating from the top skin downward. The NDE methods used in this work for mapping out these damages were: air-coupled ultrasonic scan, and imaging by computer aided tap tester. Representative results obtained from the field will be shown

  5. Activity release from the damaged spent VVER-fuel during long-term wet storage

    International Nuclear Information System (INIS)

    Slonszki, E.; Hozer, Z.; Pinter, T.; Baracska Varju, I.

    2010-01-01

    An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10 th April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO 2 mass released from the fuel into the coolant was ∼ 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

  6. Analysis of core damage frequency from internal events: Methodology guidelines: Volume 1

    International Nuclear Information System (INIS)

    Drouin, M.T.; Harper, F.T.; Camp, A.L.

    1987-09-01

    NUREG-1150 examines the risk to the public from a selected group of nuclear power plants. This report describes the methodology used to estimate the internal event core damage frequencies of four plants in support of NUREG-1150. In principle, this methodology is similar to methods used in past probabilistic risk assessments; however, based on past studies and using analysts that are experienced in these techniques, the analyses can be focused in certain areas. In this approach, only the most important systems and failure modes are modeled in detail. Further, the data and human reliability analyses are simplified, with emphasis on the most important components and human actions. Using these methods, an analysis can be completed in six to nine months using two to three full-time systems analysts and part-time personnel in other areas, such as data analysis and human reliability analysis. This is significantly faster and less costly than previous analyses and provides most of the insights that are obtained by the more costly studies. 82 refs., 35 figs., 27 tabs

  7. Measures of total stress-induced blood pressure responses are associated with vascular damage.

    Science.gov (United States)

    Nazzaro, Pietro; Seccia, Teresa; Vulpis, Vito; Schirosi, Gabriella; Serio, Gabriella; Battista, Loredana; Pirrelli, Anna

    2005-09-01

    The role of cardiovascular reactivity to study hypertension, and the assessment methods, are still controversial. We aimed to verify the association of hypertension and vascular damage with several measures of cardiovascular response. We studied 40 patients with normal-high (132 +/- 1/87 +/- 1 mm Hg) blood pressure (Group 1) and 80 untreated hypertensive subjects. Postischemic forearm vascular resistance (mFVR) served to differentiate hypertensive subjects (142 +/- 2/92 +/- 1 mm Hg v 143 +/- 2/94 +/- 2 mm Hg, P = NS) with a lower (Group 2) and higher (Group 3) hemodynamic index of vascular damage (4.8 +/- .05 v 6.3 +/- .09, P blood pressure, heart rate, forearm blood flow, and vascular resistance. Reactivity measures included: a) change from baseline, b) residualized score, c) cumulative change from baseline and residualized score, and d) total reactivity as area-under-the-curve (AUC), including changes occurring during baseline and recovery phases. The AUC of systolic blood pressure, diastolic blood pressure, and mFVR progressively increased in the groups (P AUC of SBP, DBP, and forearm blood flow and resistance demonstrated the highest (P AUC of SBP (beta = 0.634) and forearm blood flow (beta = -0.337) were predictive (P blood pressure stress response, as AUC, including baseline and recovery phases, was significantly better associated with hypertension and vascular damage than the other reactivity measures studied.

  8. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  9. Support tube of in-core instruments

    International Nuclear Information System (INIS)

    Suzumura, Takeshi; Saito, Shozo; Yasuda, Tetsuo; Shirosaki, Kiyotaka.

    1975-01-01

    Object: To permit satisfactory output measurement by preventing the bending of a in-core instrument tube within a reactor due to vibrations by means of a spring and thereby preventing mechanical damage of an adjacent fuel channel box. Structure: At a corner of a channel box of a fuel assembly, a in-core instrument tube is arranged along a channel box and has its surface provided with a plurality of removable leaf springs arranged in the direction of axis of the in-core instrument tube and each having an arcular tip. Thus, when the in-core instrument tube is inserted into the reactor, the arcular tip portions of the leaf springs are brought into plane contact with the corner of the channel box so that the in-core instrument tube is elastically supported on the channel box. Thus, there is no possibility of causing damage to the adjacent fuel channel box. (Kamimura, M.)

  10. Thermal interaction of core melt debris with the TMI-2 baffle, core-former, and lower head structures

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Tolman, E.L.

    1987-09-01

    Recent inspection of the TMI-2 core-former baffle walls (vertical), former plates (horizontal), and lower plenum has been conducted to assess potential damage to these structures. Video observations show evidence of localized melt failure of the baffle walls, whereas fiberoptics data indicate the presence of resolidified debris on the former plates. Lower plenum inspection also confirms the presence of 20 tons or more of core debris in the lower plenum. These data indicate massive core melt relocation and the potential for melt attack on vessel structural components. This report presents analyses aimed at developing an understanding of melt relocation behavior and damage progression to TMI-2 vessel components. Thermal analysis indicates melt-through of the baffle plates, but maintenance of structural integrity of the former plates and lower head. Differences in the damage of these structures is attributed largely to differences in contact time with melt debris and pressure of water. 29 refs., 17 figs., 9 tabs

  11. One-speed neutron transport in spheres with totally absorbing cores

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.

    1988-01-01

    Stationary and time-dependent transport of neutrons of one speed has been studied in spheres with totally absorbing cores. For stationary, critical reactors the number of secondaries per collision has been calculated numerically for various inner and outer radii. In the time-dependent case, the decay constant has been calculated for spherical shells of different inner radii and thicknesses. For a fixed ratio between shell thickness and inner radius, the curve of the decay constant versus shell thickness crosses the Corngold limit in the same way as the curve for a homogeneous sphere. When the ratio goes to zero the curve approaches that for an infinite slab. The behaviour is discussed in view of a new result from collision theory, viz. that the following condition must be fulfilled for a body at the point where the decay constant curve crosses the Corngold limit: the average exit distance of the neutrons is equal to the mean free path for scattering

  12. Activity release from the damaged spent VVER-fuel during long-term wet storage

    Energy Technology Data Exchange (ETDEWEB)

    Slonszki, E.; Hozer, Z. [Hungarian Academy of Sciences, KFKI Atomic Energy Research Inst., Budapest (Hungary); Pinter, T.; Baracska Varju, I. [Nuclear Power Plant Paks, Paks (Hungary)

    2010-07-01

    An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10{sup th} April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO{sub 2} mass released from the fuel into the coolant was {approx} 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

  13. TMI-2 core examination plan

    International Nuclear Information System (INIS)

    Owen, D.E.; MacDonald, P.E.; Hobbins, R.R.; Ploggr, S.A.

    1982-01-01

    The Three Mile Island (TMI-2) core examination is divided into four stages: (1) before removing the head; (2) before removing the plenum; (3) during defueling; and (4) offsite examinations. Core examinations recommended during the first three stages are primarily devoted to documenting the post-accident condition of the core. The detailed analysis of core damage structures will be performed during offsite examinations at government and commercial hot cell facilities. The primary objectives of these examinations are to enhance the understanding of the degraded core accident sequence, to develop the technical bases for reactor regulations, and to improve LWR design and operation

  14. Centennial-scale records of total organic carbon in sediment cores from the South Yellow Sea, China

    Science.gov (United States)

    Zhu, Qing; Lin, Jia; Hong, Yuehui; Yuan, Lirong; Liu, Jinzhong; Xu, Xiaoming; Wang, Jianghai

    2018-01-01

    Global carbon cycling is a significant factor that controls climate change. The centennial-scale variations in total organic carbon (TOC) contents and its sources in marginal sea sediments may reflect the influence of human activities on global climate change. In this study, two fine-grained sediment cores from the Yellow Sea Cold Water Mass of the South Yellow Sea were used to systematically determine TOC contents and stable carbon isotope ratios. These results were combined with previous data of black carbon and 210Pb dating from which we reconstructed the centennial-scale initial sequences of TOC, terrigenous TOC (TOCter) and marine autogenous TOC (TOCmar) after selecting suitable models to correct the measured TOC (TOCcor). These sequences showed that the TOCter decreased with time in the both cores while the TOCmar increased, particularly the rapid growth in core H43 since the late 1960s. According to the correlation between the Huanghe (Yellow) River discharge and the TOCcor, TOCter, or TOCmar, we found that the TOCter in the two cores mainly derived from the Huanghe River and was transported by it, and that higher Huanghe River discharge could strengthen the decomposition of TOCmar. The newly obtained initial TOC sequences provide important insights into the interaction between human activities and natural processes.

  15. 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

  16. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-01-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  17. Determination of PWR core water level using ex-core detectors signals

    International Nuclear Information System (INIS)

    Bernal, Alvaro; Abarca, Agustin; Miro, Rafael; Verdu, Gumersindo

    2013-01-01

    The core water level provides relevant neutronic and thermalhydraulic information of the reactor such as power, k eff and cooling ability; in fact, core water level monitoring could be used to predict LOCA and cooling reduction which may deal with core damage. Although different detection equipment is used to monitor several parameters such as the power, core water level monitoring is not an evident task. However, ex-core detectors can measure the fast neutrons leaking the core and several studies demonstrate the existence of a relationship between fast neutron leakage and core water level due to the shielding effect of the water. In addition, new ex-core detectors are being developed, such as silicon carbide semiconductor radiation detectors, monitoring the neutron flux with higher accuracy and in higher temperatures conditions. Therefore, a methodology to determine this relationship has been developed based on a Monte Carlo calculation using MCNP code and applying variance reduction with adjoint functions based on the adjoint flux obtained with the discrete ordinates code TORT. (author)

  18. Studies on WWER core diagnostics

    International Nuclear Information System (INIS)

    Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

    1987-01-01

    The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

  19. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  20. War damages and reconstruction of Peruca dam

    International Nuclear Information System (INIS)

    Nonveiller, E.; Sever, Z.

    1999-01-01

    The paper describes the heavy damages caused by blasting in the Peruca rockfill dam in Croatia in January 1993. Complete collapse of the dam by overtopping was prevented through quick action of the dam owner by dumping clayey gravel on the lowest sections of the dam crest and opening the bottom outlet of the reservoir, thus efficiently lowering the water level. After the damages were sufficiently established and alternatives for restoration of the dam were evaluated, it was decided to construct a diaphragm wall through the damaged core in the central dam part as the impermeable dam element and to rebuild the central clay core at the dam abutments. Reconstruction works are described

  1. Use of Added Sugars Instead of Total Sugars May Improve the Capacity of the Health Star Rating System to Discriminate between Core and Discretionary Foods.

    Science.gov (United States)

    Menday, Hannah; Neal, Bruce; Wu, Jason H Y; Crino, Michelle; Baines, Surinder; Petersen, Kristina S

    2017-12-01

    The Australian Government has introduced a voluntary front-of-package labeling system that includes total sugar in the calculation. Our aim was to determine the effect of substituting added sugars for total sugars when calculating Health Star Ratings (HSR) and identify whether use of added sugars improves the capacity to distinguish between core and discretionary food products. This study included packaged food and beverage products available in Australian supermarkets (n=3,610). The product categories included in the analyses were breakfast cereals (n=513), fruit (n=571), milk (n=309), non-alcoholic beverages (n=1,040), vegetables (n=787), and yogurt (n=390). Added sugar values were estimated for each product using a validated method. HSRs were then estimated for every product according to the established method using total sugar, and then by substituting added sugar for total sugar. The scoring system was not modified when added sugar was used in place of total sugar in the HSR calculation. Products were classified as core or discretionary based on the Australian Dietary Guidelines. To investigate whether use of added sugar in the HSR algorithm improved the distinction between core and discretionary products as defined by the Australian Dietary Guidelines, the proportion of core products that received an HSR of ≥3.5 stars and the proportion of discretionary products that received an HSR of added sugars were determined. There were 2,263 core and 1,347 discretionary foods; 1,684 of 3,610 (47%) products contained added sugar (median 8.4 g/100 g, interquartile range=5.0 to 12.2 g). When the HSR was calculated with added sugar instead of total sugar, an additional 166 (7.3%) core products received an HSR of ≥3.5 stars and 103 (7.6%) discretionary products received a rating of ≥3.5 stars. The odds of correctly identifying a product as core vs discretionary were increased by 61% (odds ratio 1.61, 95% CI 1.26 to 2.06; Padded compared to total sugars. In the six

  2. Pulsed laser damage to optical fibers

    International Nuclear Information System (INIS)

    Allison, S.W.; Gillies, G.T.; Magnuson, D.W.; Pagano, T.S.

    1985-01-01

    This paper describes some observations of pulsed laser damage to optical fibers with emphasis on a damage mode characterized as a linear fracture along the outer core of a fiber. Damage threshold data are presented which illustrate the effects of the focusing lens, end-surface preparation, and type of fiber. An explanation based on fiber-beam misalignment is given and is illustrated by a simple experiment and ray trace

  3. Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Vickers, Lisa

    2003-01-01

    The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

  4. Repair pathways independent of the Fanconi anemia nuclear core complex play a predominant role in mitigating formaldehyde-induced DNA damage

    International Nuclear Information System (INIS)

    Noda, Taichi; Takahashi, Akihisa; Kondo, Natsuko; Mori, Eiichiro; Okamoto, Noritomo; Nakagawa, Yosuke; Ohnishi, Ken; Zdzienicka, Malgorzata Z.; Thompson, Larry H.; Helleday, Thomas; Asada, Hideo

    2011-01-01

    The role of the Fanconi anemia (FA) repair pathway for DNA damage induced by formaldehyde was examined in the work described here. The following cell types were used: mouse embryonic fibroblast cell lines FANCA -/- , FANCC -/- , FANCA -/- C -/- , FANCD2 -/- and their parental cells, the Chinese hamster cell lines FANCD1 mutant (mt), FANCGmt, their revertant cells, and the corresponding wild-type (wt) cells. Cell survival rates were determined with colony formation assays after formaldehyde treatment. DNA double strand breaks (DSBs) were detected with an immunocytochemical γH2AX-staining assay. Although the sensitivity of FANCA -/- , FANCC -/- and FANCA -/- C -/- cells to formaldehyde was comparable to that of proficient cells, FANCD1mt, FANCGmt and FANCD2 -/- cells were more sensitive to formaldehyde than the corresponding proficient cells. It was found that homologous recombination (HR) repair was induced by formaldehyde. In addition, γH2AX foci in FANCD1mt cells persisted for longer times than in FANCD1wt cells. These findings suggest that formaldehyde-induced DSBs are repaired by HR through the FA repair pathway which is independent of the FA nuclear core complex. -- Research highlights: → We examined to clarify the repair pathways of formaldehyde-induced DNA damage. Formaldehyde induces DNA double strand breaks (DSBs). → DSBs are repaired through the Fanconi anemia (FA) repair pathway. → This pathway is independent of the FA nuclear core complex. → We also found that homologous recombination repair was induced by formaldehyde.

  5. Tailoring Sandwich Face/Core Interfaces for Improved Damage Tolerance

    DEFF Research Database (Denmark)

    Lundsgaard-Larsen, Christian; Berggreen, Christian; Carlsson, Leif A.

    2010-01-01

    Various modifications of the face/core interface in foam core sandwich specimens are examined in a series of two papers. This paper constitutes part I and describes the finite element analysis of a sandwich test specimen, i.e. a DCB specimen loaded by uneven bending moments (DCB-UBM). Using...... this test almost any mode-mixity between pure mode I and mode II can be obtained. A cohesive zone model of the mixed mode fracture process involving large-scale bridging is developed. Results from the analysis are used in Part II, which describes methods and results of a series of experiments....

  6. Changes in nuclear protein acetylation in u. v. -damaged human cells

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, B.; Smerdon, M.J.

    1986-07-01

    We have investigated the levels of nuclear protein acetylation in u.v.-irradiated human fibroblasts. We measured the levels of acetylation in total acid-soluble nuclear proteins and observed two distinct differences between the irradiated and unirradiated (control) cells. Immediately after irradiation, there is a wave of protein hyperacetylation (i.e. a total acetylation level greater than that of unirradiated cells) that lasts for 2-6 h depending on the experimental conditions. This hyperacetylation phase is then followed by a hypoacetylation phase, lasting for many hours, and the total level of acetylation does not return to that of control cells until 24-72 h after u.v. damage. Both the magnitude and duration of each phase is dependent on the dose of u.v. light used. The wave of hyperacetylation is more pronounced at low u.v. doses (i.e. less than 5 J/m2), while the wave of hypoacetylation is more pronounced at higher u.v. doses (greater than or equal to 8 J/m2). Furthermore, the duration of each phase is prolonged when cells are exposed to 2 mM hydroxyurea. Examination of the acetylation levels of the individual nuclear proteins indicated that acetylation of the core histones follows the same pattern observed for the total acid-soluble protein fractions. Furthermore, these were the only major proteins in the total acid-soluble fraction observed to undergo the early, rapid hyperacetylation immediately following u.v. damage. Acetylation of histone H1 was negligible in both damaged and control cells, while three prominent non-histone proteins were acetylated only after long labeling times (greater than 4 h) in each case, gradually becoming hyperacetylated in the u.v.-damaged cells. These results raise the possibility that a causal relationship exists between nuclear protein acetylation and nucleotide excision repair of DNA in human cells.

  7. Changes in nuclear protein acetylation in u.v.-damaged human cells

    International Nuclear Information System (INIS)

    Ramanathan, B.; Smerdon, M.J.

    1986-01-01

    The levels of nuclear protein acetylation in u.v.-irradiated human fibroblasts have been investigated. Initially, we measured the levels of acetylation in total acid-soluble nuclear proteins and observed two distinct differences between the irradiated and unirradiated (control) cells. Immediately after irradiation, there is a 'wave' of protein hyperacetylation that lasts for 2-6 h, followed by a hypoacetylation phase, lasting for many hours, and the total level of acetylation does not return to that of control cells until 24-72 h after u.v. damage. Both the magnitude and duration of each phase is dependent on the dose of u.v. light used. The wave of hyperacetylation is more pronounced at low u.v. doses, while the wave of hypoacetylation is more pronounced at higher u.v. doses. Furthermore, the duration of each phase is prolonged when cells are exposed to 2 mM hydroxyurea, an agent which retards the rate of excision repair at u.v.-damaged sites. Examinations of the acetylation levels of the individual nuclear proteins indicated that acetylation of the core histones follows the same pattern observed for the total acid-soluble protein fractions. Furthermore, these were the only major proteins in the total acid-soluble fraction observed to undergo the early, rapid hyperacetylation immediately following u.v. damage. These results raise the possibility that a causal relationship exists between nuclear protein acetylation and nucleotide excision repair of DNA in human cells. (author)

  8. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  9. Comparative study of radiation damage accumulation in Cu and Fe

    International Nuclear Information System (INIS)

    Caturla, M.J.; Soneda, N.; Alonso, E.; Wirth, B.D.; Diaz de la Rubia, T.; Perlado, J.M.

    2000-01-01

    Bcc and fcc metals exhibit significant differences in behavior when exposed to neutron or heavy ion irradiation. Transmission electron microscopy (TEM) observations reveal that damage in the form of stacking fault tetrahedra (SFT) is visible in copper irradiated to very low doses, but that no damage is visible in iron irradiated to the same total dose. In order to understand and quantify this difference in behavior, we have simulated damage production and accumulation in fcc Cu and bcc Fe. We use 20 keV primary knock-on atoms (PKAs) at a homologous temperature of 0.25 of the melting point. The primary damage state was calculated using molecular dynamics (MD) with empirical, embedded-atom interatomic potentials. Damage accumulation was modeled using a kinetic Monte Carlo (kMC) algorithm to follow the evolution of all defects produced in the cascades. The diffusivities and binding energies of defects are input data for this simulation and were either extracted from experiments, the literature, or calculated using MD. MD simulations reveal that vacancy clusters are produced within the cascade core in the case of copper. In iron, most of the vacancies do not cluster during cooling of the cascade core and are available for diffusion. In addition, self-interstitial atom (SIA) clusters are produced in copper cascades but those observed in iron are smaller in number and size. The combined MD/kMC simulations reveal that the visible cluster densities obtained as a function of dose are at least one order of magnitude lower in Fe than in Cu. We compare the results with experimental measurements of cluster density and find excellent agreement between the simulations and experiments when small interstitial clusters are considered to be mobile as suggested by recent MD simulations

  10. USNRC severe core damage assessment program

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, J E [EG and G Idaho, Inc., Idaho Falls (USA); Johnston, W V; Kelber, C N [Nuclear Regulatory Commission, Washington, DC (USA)

    1981-01-01

    The accident at the Three Mile Island nuclear power station has significantly altered the perception of the importance of beyond-design-basis accidents in licensing and safety reviews of light-water reactors in the USA. Increased consideration will be given by the United States Nuclear Regulatory Commission to low-probability, high-risk core melt accidents in future licensing proceedings. To this end, the USNRC is mounting experimental and analytic methods development programs to provide the technical basis for future LWR design and licensing criteria related to class-9 accidents. The scope, objectives, and content of five major new programs addressing safety and licensing issues for beyond-design-basis accidents are reviewed and the rationale and logic for formulation of the programs is discussed.

  11. Compendium of Single Event Effects, Total Ionizing Dose, and Displacement Damage for Candidate Spacecraft Electronics for NASA

    Science.gov (United States)

    LaBel, Kenneth A.; OBryan, Martha V.; Chen, Dakai; Campola, Michael J.; Casey, Megan C.; Pellish, Jonathan A.; Lauenstein, Jean-Marie; Wilcox, Edward P.; Topper, Alyson D.; Ladbury, Raymond L.; hide

    2014-01-01

    We present results and analysis investigating the effects of radiation on a variety of candidate spacecraft electronics to proton and heavy ion induced single event effects (SEE), proton-induced displacement damage (DD), and total ionizing dose (TID). Introduction: This paper is a summary of test results.NASA spacecraft are subjected to a harsh space environment that includes exposure to various types of ionizing radiation. The performance of electronic devices in a space radiation environment is often limited by its susceptibility to single event effects (SEE), total ionizing dose (TID), and displacement damage (DD). Ground-based testing is used to evaluate candidate spacecraft electronics to determine risk to spaceflight applications. Interpreting the results of radiation testing of complex devices is quite difficult. Given the rapidly changing nature of technology, radiation test data are most often application-specific and adequate understanding of the test conditions is critical. Studies discussed herein were undertaken to establish the application-specific sensitivities of candidate spacecraft and emerging electronic devices to single-event upset (SEU), single-event latchup (SEL), single-event gate rupture (SEGR), single-event burnout (SEB), single-event transient (SET), TID, enhanced low dose rate sensitivity (ELDRS), and DD effects.

  12. Haloperidol-loaded lipid-core polymeric nanocapsules reduce DNA damage in blood and oxidative stress in liver and kidneys of rats

    International Nuclear Information System (INIS)

    Roversi, Katiane; Benvegnú, Dalila M.; Roversi, Karine; Trevizol, Fabíola; Vey, Luciana T.; Elias, Fabiana; Fracasso, Rafael

    2015-01-01

    Haloperidol (HP) nanoencapsulation improves therapeutic efficacy, prolongs the drug action time, and reduces its motor side effects. However, in a view of HP toxicity in organs like liver and kidneys in addition to the lack of knowledge regarding the toxicity of polymeric nanocapsules, our aim was to verify the influence of HP-nanoformulation on toxicity and oxidative stress markers in the liver and kidneys of rats, also observing the damage caused in the blood. For such, 28 adult male Wistar rats were designated in four experimental groups (n = 7) and treated with vehicle (C group), free haloperidol suspension (FH group), blank nanocapsules suspension (B-Nc group), and haloperidol-loaded lipid-core nanocapsules suspension (H-Nc group). The nanocapsules formulation presented the size of approximately 250 nm. All suspensions were administered to the animals (0.5 mg/kg/day-i.p.) for a period of 28 days. Our results showed that FH caused damage in the liver, evidenced by increased lipid peroxidation, plasma levels of aspartate aminotransferase, and alanine aminotransferase, as well as decreased cellular integrity and vitamin C levels. In kidneys, FH treatment caused damage to a lesser extent, observed by decreased activity of δ-aminolevulinate dehydratase (ALA-D) and levels of VIT C. In addition, FH treatment was also related to a higher DNA damage index in blood. On the other hand, animals treated with H-Nc and B-Nc did not show damage in liver, kidneys, and DNA. Our study indicates that the nanoencapsulation of haloperidol was able to prevent the sub-chronic toxicity commonly observed in liver, kidneys, and DNA, thus reflecting a pharmacological superiority in relation to free drug

  13. Haloperidol-loaded lipid-core polymeric nanocapsules reduce DNA damage in blood and oxidative stress in liver and kidneys of rats

    Science.gov (United States)

    Roversi, Katiane; Benvegnú, Dalila M.; Roversi, Karine; Trevizol, Fabíola; Vey, Luciana T.; Elias, Fabiana; Fracasso, Rafael; Motta, Mariana H.; Ribeiro, Roseane F.; dos S. Hausen, Bruna; Moresco, Rafael N.; Garcia, Solange C.; da Silva, Cristiane B.; Burger, Marilise E.

    2015-04-01

    Haloperidol (HP) nanoencapsulation improves therapeutic efficacy, prolongs the drug action time, and reduces its motor side effects. However, in a view of HP toxicity in organs like liver and kidneys in addition to the lack of knowledge regarding the toxicity of polymeric nanocapsules, our aim was to verify the influence of HP-nanoformulation on toxicity and oxidative stress markers in the liver and kidneys of rats, also observing the damage caused in the blood. For such, 28 adult male Wistar rats were designated in four experimental groups ( n = 7) and treated with vehicle (C group), free haloperidol suspension (FH group), blank nanocapsules suspension (B-Nc group), and haloperidol-loaded lipid-core nanocapsules suspension (H-Nc group). The nanocapsules formulation presented the size of approximately 250 nm. All suspensions were administered to the animals (0.5 mg/kg/day-i.p.) for a period of 28 days. Our results showed that FH caused damage in the liver, evidenced by increased lipid peroxidation, plasma levels of aspartate aminotransferase, and alanine aminotransferase, as well as decreased cellular integrity and vitamin C levels. In kidneys, FH treatment caused damage to a lesser extent, observed by decreased activity of δ-aminolevulinate dehydratase (ALA-D) and levels of VIT C. In addition, FH treatment was also related to a higher DNA damage index in blood. On the other hand, animals treated with H-Nc and B-Nc did not show damage in liver, kidneys, and DNA. Our study indicates that the nanoencapsulation of haloperidol was able to prevent the sub-chronic toxicity commonly observed in liver, kidneys, and DNA, thus reflecting a pharmacological superiority in relation to free drug.

  14. Haloperidol-loaded lipid-core polymeric nanocapsules reduce DNA damage in blood and oxidative stress in liver and kidneys of rats

    Energy Technology Data Exchange (ETDEWEB)

    Roversi, Katiane, E-mail: katianeroversi@gmail.com [Universidade Federal de Santa Maria, Programa de Pós-Graduação em Farmacologia (Brazil); Benvegnú, Dalila M., E-mail: dalilabenvegnu@yahoo.com.br [Universidade Federal da Fronteira Sul (UFFS), Bioquímica e Farmacologia (Brazil); Roversi, Karine, E-mail: karineroversi-@hotmail.com [Universidade Federal de Santa Maria (UFSM), Departamento de Fisiologia e Farmacologia, Centro de Ciências da Saúde (Brazil); Trevizol, Fabíola, E-mail: fatrevizol@yahoo.com.br [Universidade Federal de Santa Maria, Programa de Pós-Graduação em Farmacologia (Brazil); Vey, Luciana T., E-mail: luciana.taschetto@hotmail.com [Universidade Federal de Santa Maria (UFSM), Departamento de Fisiologia e Farmacologia, Centro de Ciências da Saúde (Brazil); Elias, Fabiana, E-mail: fabiana.elias@uffs.edu.br [Universidade Federal da Fronteira Sul (UFFS), Bioquímica e Farmacologia (Brazil); Fracasso, Rafael, E-mail: rafael.fra@hotmail.com [Universidade Federal do Rio Grande do Sul, Programa de Pós-Graduação em Ciências Farmacêuticas (Brazil); and others

    2015-04-15

    Haloperidol (HP) nanoencapsulation improves therapeutic efficacy, prolongs the drug action time, and reduces its motor side effects. However, in a view of HP toxicity in organs like liver and kidneys in addition to the lack of knowledge regarding the toxicity of polymeric nanocapsules, our aim was to verify the influence of HP-nanoformulation on toxicity and oxidative stress markers in the liver and kidneys of rats, also observing the damage caused in the blood. For such, 28 adult male Wistar rats were designated in four experimental groups (n = 7) and treated with vehicle (C group), free haloperidol suspension (FH group), blank nanocapsules suspension (B-Nc group), and haloperidol-loaded lipid-core nanocapsules suspension (H-Nc group). The nanocapsules formulation presented the size of approximately 250 nm. All suspensions were administered to the animals (0.5 mg/kg/day-i.p.) for a period of 28 days. Our results showed that FH caused damage in the liver, evidenced by increased lipid peroxidation, plasma levels of aspartate aminotransferase, and alanine aminotransferase, as well as decreased cellular integrity and vitamin C levels. In kidneys, FH treatment caused damage to a lesser extent, observed by decreased activity of δ-aminolevulinate dehydratase (ALA-D) and levels of VIT C. In addition, FH treatment was also related to a higher DNA damage index in blood. On the other hand, animals treated with H-Nc and B-Nc did not show damage in liver, kidneys, and DNA. Our study indicates that the nanoencapsulation of haloperidol was able to prevent the sub-chronic toxicity commonly observed in liver, kidneys, and DNA, thus reflecting a pharmacological superiority in relation to free drug.

  15. TMI-2 core bore acquisition summary report

    International Nuclear Information System (INIS)

    Tolman, E.L.; Smith, R.P.; Martin, M.R.; McCardell, R.K.; Broughton, J.M.

    1986-09-01

    Core bore samples were obtained from the severely damaged TMI-2 core during July and August, 1986. A description of the TMI-2 core bore drilling unit used to obtain samples; a summary and discussion of the data from the ten core bore segments which were obtained; and the initial results of analysis and evaluation of these data are presented in this report. The impact of the major findings relative to our understanding of the accident scenario is also discussed

  16. Effect of vitamin E and C supplementation on oxidative damage and total antioxidant capacity in lead-exposed workers.

    Science.gov (United States)

    Rendón-Ramírez, Adela-Leonor; Maldonado-Vega, María; Quintanar-Escorza, Martha-Angelica; Hernández, Gerardo; Arévalo-Rivas, Bertha-Isabel; Zentella-Dehesa, Alejandro; Calderón-Salinas, José-Víctor

    2014-01-01

    The molecular response of the antioxidant system and the effects of antioxidant supplementation against oxidative insult in lead-exposed workers has not been sufficiently studied. In this work, antioxidants (vitamin E 400 IU+vitamin C 1g/daily) were supplemented for one year to 15 workers exposed to lead (73 μg of lead/dl of blood) and the results were compared with those on 19 non-lead exposed workers (6.7 μg of lead/dl). Lead intoxication was accompanied by a high oxidative damage and an increment in the erythrocyte antioxidant response due to increased activity of catalase and superoxide dismutase. Antioxidant supplementations decreased significantly the oxidative damage as well as the total antioxidant capacity induced by lead intoxication with reduction of the antioxidant enzyme activities. We conclude that antioxidant supplementation is effective in reducing oxidative damage and induces modifications in the physiopathological status of the antioxidant response in lead-exposed workers. Copyright © 2013 Elsevier B.V. All rights reserved.

  17. Comparing damage on retrieved total elbow replacement bushings with lab worn specimens subjected to varied loading conditions.

    Science.gov (United States)

    Willing, Ryan

    2018-01-09

    Complication rates following total elbow replacement (TER) with conventional implants are relatively high due to mechanical failure involving the UHMWPE bushings. Unfortunately, there are no standardized pre-clinical durability testing protocols for assessing the durability of TER components. This study examines the damage observed on retrieved humeral bushings, and then uses in vitro durability testing with two different loading protocols to compare resulting damage. Damage on 25 pairs of retrieved humeral bushings was characterized using micro-computed tomographic imaging techniques. The damage was compared with that of in vitro test specimens which were subjected to 200 K cycles of either high joint reaction force (high JRF) or high varus moment (high VM) loading. Material removal (mass loss) from bushing components was measured using gravimetric techniques. Thinning was less for retrieved bushings which were still assembled in their humeral component, versus bushings which were loose (0.3 ± 0.3 mm vs. 0.6 ± 0.3 mm, p = 0.02). Comparing in vitro test specimens, thinning due to high VM loading was 0.9 ± 0.3 mm, versus 0.2 ± 0.0 mm for high JRF loading (p = 0.08); however, the actual material removal rates from the humeral bushings were not different between the two protocols (48 ± 5 mm 3 /Mc vs. 43 ± 2 mm 3 /Mc, p = 1). Neither loading protocol could produce damage patterns fully representative of the spectrum of damage patterns observed on clinical retrievals. Pre-clinical testing should employ multiple loading protocols to characterize implant performance under a broader spectrum of usage. © 2018 Orthopaedic Research Society. Published by Wiley Periodicals, Inc. J Orthop Res. © 2018 Orthopaedic Research Society. Published by Wiley Periodicals, Inc.

  18. Repair pathways independent of the Fanconi anemia nuclear core complex play a predominant role in mitigating formaldehyde-induced DNA damage

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Taichi [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Department of Dermatology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Takahashi, Akihisa [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Kondo, Natsuko [Particle Radiation Oncology Research Center, Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Mori, Eiichiro [Department of Biology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Okamoto, Noritomo [Department of Otorhinolaryngology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Nakagawa, Yosuke [Department of Oral and Maxillofacial Surgery, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); Ohnishi, Ken [Department of Biology, Ibaraki Prefectual University of Health Sciences, 4669-2 Ami, Ami-mati, Inasiki-gun, Ibaraki 300-0394 (Japan); Zdzienicka, Malgorzata Z. [Department of Molecular Cell Genetics, Collegium Medicum in Bydgoszcz, Nicolaus-Copernicus-University in Torun, ul. Sklodowskiej-Curie 9, 85-094 Bydgoszcz (Poland); Thompson, Larry H. [Biosciences and Biotechnology Division, L452, Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Helleday, Thomas [Gray Institute for Radiation Oncology and Biology, University of Oxford, Old Road Campus Research Building, Off Roosevelt Drive, Oxford, OX3 7DQ (United Kingdom); Department of Genetics, Microbiology and Toxicology Stockholm University, SE-106 91 Stockholm (Sweden); Asada, Hideo [Department of Dermatology, School of Medicine, Nara Medical University, 840 Shijo-cho, Kashihara, Nara 634-8521 (Japan); and others

    2011-01-07

    The role of the Fanconi anemia (FA) repair pathway for DNA damage induced by formaldehyde was examined in the work described here. The following cell types were used: mouse embryonic fibroblast cell lines FANCA{sup -/-}, FANCC{sup -/-}, FANCA{sup -/-}C{sup -/-}, FANCD2{sup -/-} and their parental cells, the Chinese hamster cell lines FANCD1 mutant (mt), FANCGmt, their revertant cells, and the corresponding wild-type (wt) cells. Cell survival rates were determined with colony formation assays after formaldehyde treatment. DNA double strand breaks (DSBs) were detected with an immunocytochemical {gamma}H2AX-staining assay. Although the sensitivity of FANCA{sup -/-}, FANCC{sup -/-} and FANCA{sup -/-}C{sup -/-} cells to formaldehyde was comparable to that of proficient cells, FANCD1mt, FANCGmt and FANCD2{sup -/-} cells were more sensitive to formaldehyde than the corresponding proficient cells. It was found that homologous recombination (HR) repair was induced by formaldehyde. In addition, {gamma}H2AX foci in FANCD1mt cells persisted for longer times than in FANCD1wt cells. These findings suggest that formaldehyde-induced DSBs are repaired by HR through the FA repair pathway which is independent of the FA nuclear core complex. -- Research highlights: {yields} We examined to clarify the repair pathways of formaldehyde-induced DNA damage. Formaldehyde induces DNA double strand breaks (DSBs). {yields} DSBs are repaired through the Fanconi anemia (FA) repair pathway. {yields} This pathway is independent of the FA nuclear core complex. {yields} We also found that homologous recombination repair was induced by formaldehyde.

  19. Radiation damage studies of nuclear structural materials

    International Nuclear Information System (INIS)

    Barat, P.

    2012-01-01

    Maximum utilization of fuel in nuclear reactors is one of the important aspects for operating them economically. The main hindrance to achieve this higher burnups of nuclear fuel for the nuclear reactors is the possibility of the failure of the metallic core components during their operation. Thus, the study of the cause of the possibility of failure of these metallic structural materials of nuclear reactors during full power operation due to radiation damage, suffered inside the reactor core, is an important field of studies bearing the basic to industrial scientific views.The variation of the microstructure of the metallic core components of the nuclear reactors due to radiation damage causes enormous variation in the structure and mechanical properties. A firm understanding of this variation of the mechanical properties with the variation of microstructure will serve as a guide for creating new, more radiation-tolerant materials. In our centre we have irradiated structural materials of Indian nuclear reactors by charged particles from accelerator to generate radiation damage and studied the some aspects of the variation of microstructure by X-ray diffraction studies. Results achieved in this regards, will be presented. (author)

  20. Muscle Damage After Total Hip Arthroplasty Through the Direct Anterior Approach for Developmental Dysplasia of the Hip.

    Science.gov (United States)

    Kawasaki, Masashi; Hasegawa, Yukiharu; Okura, Toshiaki; Ochiai, Satoshi; Fujibayashi, Takayoshi

    2017-08-01

    Total hip arthroplasty (THA) through the direct anterior approach (DAA) is known to cause less muscle damage than other surgical approaches. However, more complex primary cases, such as developmental dysplasia of the hip (DDH), might often cause muscle damage. The objective of the present study was to clarify the muscle damage observed 1 year after THA through the DAA for DDH using magnetic resonance imaging. We prospectively compared the muscle cross-sectional area (M-CSA) and fatty atrophy (FA) in muscles by magnetic resonance imaging and the Harris hip score before and at 1-year follow-up after THA through the DAA in 3 groups: 37 patients with Crowe group 1 DDH (D1), 13 patients with Crowe group 2 and 3 DDH (D2 + 3), and 12 patients with osteonecrosis as a control. THA through the DAA for D1 displayed significantly decreased M-CSA and significantly increased FA in the gluteus minimus (Gmini), the tensor fasciae latae (TFL), and the obturator internus (OI). Patients with D2 + 3 group did not have decreased M-CSA in the TFL or increased FA in the Gmini. Postoperatively, a significant negative correlation was observed between the M-CSA and FA for the OI in patients with D1 and D2 + 3. THA through the DAA for DDH caused the damage in the Gmini, the TFL, and the OI; severe damage was observed in the OI, showing increased FA with decreased M-CSA in patients with both D1 and D2 + 3. Copyright © 2017 Elsevier Inc. All rights reserved.

  1. Validation of the mortality prediction equation for damage control ...

    African Journals Online (AJOL)

    , preoperative lowest pH and lowest core body temperature to derive an equation for the purpose of predicting mortality in damage control surgery. It was shown to reliably predict death despite damage control surgery. The equation derivation ...

  2. Does core mobility of lumbar total disc arthroplasty influence sagittal and frontal intervertebral displacement? Radiologic comparison with fixed-core prosthesis

    Science.gov (United States)

    Delécrin, Joël; Allain, Jérôme; Beaurain, Jacques; Steib, Jean-Paul; Chataigner, Hervé; Aubourg, Lucie; Huppert, Jean; Ameil, Marc; Nguyen, Jean-Michel

    2009-01-01

    Background An artificial disc prosthesis is thought to restore segmental motion in the lumbar spine. However, it is reported that disc prosthesis can increase the intervertebral translation (VT). The concept of the mobile-core prosthesis is to mimic the kinematic effects of the migration of the natural nucleus and therefore core mobility should minimize the VT. This study explored the hypothesis that core translation should influence VT and that a mobile core prosthesis may facilitate physiological motion. Methods Vertebral translation (measured with a new method presented here), core translation, range of motion (ROM), and distribution of flexion-extension were measured on flexion-extension, neutral standing, and lateral bending films in 89 patients (63 mobile-core [M]; 33 fixed-core [F]). Results At L4-5 levels the VT with M was lower than with F and similar to the VT of untreated levels. At L5-S1 levels the VT with M was lower than with F but was significantly different compared to untreated levels. At M levels a strong correlation was found between VT and core translation; the VT decreases as the core translation increases. At F levels the VT increases as the ROM increases. No significant difference was found between the ROM of untreated levels and levels implanted with either M or F. Regarding the mobility distribution with M and F we observed a deficit in extension at L5-S1 levels and a similar distribution at L4-5 levels compared to untreated levels. Conclusion The intervertebral mobility was different between M and F. The M at L4-5 levels succeeded to replicate mobility similar to L4-5 untreated levels. The M at L5-S1 succeeded in ROM, but failed regarding VT and mobility distribution. Nevertheless M minimized VT at L5-S1 levels. The F increased VT at both L4-5 and L5-S1. Clinical Relevance This study validates the concept that the core translation of an artificial lumbar disc prosthesis minimizes the VT. PMID:25802632

  3. PBF severe fuel damage program: results and comparison to analysis

    International Nuclear Information System (INIS)

    McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  4. PBF Severe Fuel-Damage Program: results and comparison to analysis

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  5. The contribution to site core damage frequency from independent occurrences of initiators in two or more units: How low is it?

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-San; Park, Jin Hee; Lim, Ho Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Stutzke estimated the site risk by summing the contribution from common cause initiators and the contribution from single-unit initiators. He considered some kinds of multi-unit accident sequences caused by single-unit initiators. However, the contribution from independent occurrences of initiators in two or more units at a site was not taken into account. The purpose of this study is to estimate the contribution to site core damage frequency (CDF) from simultaneous occurrences of independent initiators in two or more units at the same site. Some assumptions and methods used in this analysis are firstly described, and the results and conclusions of the analysis are described. In this study, the contribution to site core damage frequency (CDF) from simultaneous occurrences of independent initiators in two or more units at the same site was estimated. A Korean six-unit site was selected as the reference site and the at-power internal events Level 1 PSA model for an OPR1000 unit at the reference site was used as the base model, and was modified to deal with some major dependencies between units at the site. Specifically, the availability of the AAC D/G, dependencies between offsite power recovery actions in different unis, and inter-unit CCF modeling for risk-significant components such as diesel generators were taken into account. As a result, the sum of dual-unit CDF due to independent occurrences of initiators in two units at the reference site was estimated to be sufficiently low to be neglected.

  6. IPE Data Base: Plant design, core damage frequency and containment performance information

    International Nuclear Information System (INIS)

    Lehner, J.; Lin, C.C.; Pratt, W.T.; Su, T.; Danziger, L.

    1995-01-01

    This data base stores data obtained from the Individual Plant Examinations (IPEs) which licensees of nuclear power plants have conducted in response to NRC's Generic Letter GL88-20. The IPE Data Base is a collection of linked files which store information about plant design, core damage frequency, and containment performance in a uniform, structured way. The information contined in the various files is based on data contained in the IPE submittals. The information extracted from the submittals and entered into the IPE Data Base can be maniulated so that queries regarding individual or groups of plants can be answered using the IPE Data Base. The IPE Data Base supports detailed inquiries into the characteristics of individual plants or classes of plants. Progress has been made on the IPE Data Base and it is largely complete. Recent focus has been the development of a user friendly version which is menu driven and allows the user to ask queries of varying complexity easily, without the need to become familiar with particular data base formats or conventions such as those of DBase IV or Microsoft Access. The user can obtain the information he desired by quickly moving through a series of on-screen menus and ''clicking'' on appropriate choices. In this way even a first time user can benefit from the large amount of information stored in the IPE Data Base without the need of a learning period

  7. Quality assurance in the removal and transport of the TMI-2 [Three Mile Island Unit 2] core

    International Nuclear Information System (INIS)

    Hayes, G.R.; Marsden, J.F.

    1988-01-01

    The March 1979 accident at Three Mile Island Unit 2 (TMI-2) damaged the core of the reactor. One of the major cleanup activities involves removal of the damaged core from the reactor and transporting it from the TMI-2 site near Middletown, Pennsylvania, to the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Removal and transport of the damaged core necessitated the development of much specialized equipment. This paper focuses on the role quality assurance (QA) played in the design, fabrication, acceptance, and use of three important pieces of core debris removal and transportation equipment: (1) the core boring machine, (2) the fuel debris canisters, (3) the NuPac 125-B rail cask and handling equipment

  8. A chronology of the PY608E-PC sediment core (Lake Pumoyum Co, southern Tibetan Plateau) based on radiocarbon dating of total organic carbon

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Takahiro, E-mail: twatanabe@geo.kankyo.tohoku.ac.jp [Graduate School of Environmental Studies, Tohoku University, 6-6-20 Aramaki Aza Aoba, Aoba-ku, Sendai 980-8579 (Japan); Graduate School of Science, Tohoku University, 6-3 Aramaki Aza Aoba, Aoba-ku, Sendai 980-8578 (Japan); Nakamura, Toshio [Center for Chronological Research, Nagoya University, Furo-cho, Chikusa, Nagoya 464-8601 (Japan); Matsunaka, Tetsuya [School of Marine Science and Technology, Tokai University, 3-20-1 Orido, Shimizu, Shizuoka 424-0902 (Japan); Nara, Fumiko Watanabe [Graduate School of Science, Tohoku University, 6-3 Aramaki Aza Aoba, Aoba-ku, Sendai 980-8578 (Japan); Zhu Liping; Wang Junbo [Institute of Tibetan Plateau Research, Chinese Academy of Science, No. 18 Shuangqing Road, Haidian District, Beijing 100085 (China); Kakegawa, Takeshi [Graduate School of Science, Tohoku University, 6-3 Aramaki Aza Aoba, Aoba-ku, Sendai 980-8578 (Japan); Nishimura, Mitsugu [School of Marine Science and Technology, Tokai University, 3-20-1 Orido, Shimizu, Shizuoka 424-0902 (Japan)

    2013-01-15

    Paleoclimatic records from the Tibetan Plateau provide important clues for understanding the Asian monsoon and Asian climate systems. To reconstruct climatic and environmental changes in the southern Tibetan Plateau, a 3.77-m-long sediment core (PY608E-PC) was taken from the southeastern part of Lake Pumoyum Co in August 2006. Because terrestrial plant residues are extremely rare in this core, we performed radiocarbon dating on the total organic carbon fraction. We also estimated the old carbon effect and radiocarbon reservoir age of the total organic carbon fraction. Using these estimates, we propose a new radiocarbon chronology for past climatic changes from ca. 12,500 to 700 cal BP. The linear sedimentation rate of the core was founded to be constant at 32.0 cm/kyr, indicating stable sedimentation conditions in Lake Pumoyum Co from the period of the Younger Dryas to the Holocene.

  9. A chronology of the PY608E–PC sediment core (Lake Pumoyum Co, southern Tibetan Plateau) based on radiocarbon dating of total organic carbon

    International Nuclear Information System (INIS)

    Watanabe, Takahiro; Nakamura, Toshio; Matsunaka, Tetsuya; Nara, Fumiko Watanabe; Zhu Liping; Wang Junbo; Kakegawa, Takeshi; Nishimura, Mitsugu

    2013-01-01

    Paleoclimatic records from the Tibetan Plateau provide important clues for understanding the Asian monsoon and Asian climate systems. To reconstruct climatic and environmental changes in the southern Tibetan Plateau, a 3.77-m-long sediment core (PY608E–PC) was taken from the southeastern part of Lake Pumoyum Co in August 2006. Because terrestrial plant residues are extremely rare in this core, we performed radiocarbon dating on the total organic carbon fraction. We also estimated the old carbon effect and radiocarbon reservoir age of the total organic carbon fraction. Using these estimates, we propose a new radiocarbon chronology for past climatic changes from ca. 12,500 to 700 cal BP. The linear sedimentation rate of the core was founded to be constant at 32.0 cm/kyr, indicating stable sedimentation conditions in Lake Pumoyum Co from the period of the Younger Dryas to the Holocene.

  10. Improvement of optical damage in specialty fiber at 266 nm wavelength

    Science.gov (United States)

    Tobisch, T.; Ohlmeyer, H.; Zimmermann, H.; Prein, S.; Kirchhof, J.; Unger, S.; Belz, M.; Klein, K.-F.

    2014-02-01

    Improved multimode UV-fibers with core diameters ranging from 70 to 600 μm diameter have been manufactured based on novel preform modifications and fiber processing techniques. Only E'-centers at 214 nm and NBOHC at 260 nm are generated in these fibers. A new generation of inexpensive laser-systems have entered the market and generated a multitude of new and attractive applications in the bio-life science, chemical and material processing field. However, for example pulsed 355 nm Nd:YAG lasers generate significant UV-damages in commercially available fibers. For lower wavelengths, no results on suitable multi-mode or low-mode fibers with high UV resistance at 266 nm wavelength (pulsed 4th harmonic Nd:YAG laser) have been published. In this report, double-clad fibers with 70 μm or 100 μm core diameter and a large claddingto- core ratio will be recommended. Laser-induced UV-damages will be compared between these new fiber type and traditional UV fibers with similar core sizes. Finally, experimental results will be cross compared against broadband cw deuterium lamp damage standards.

  11. An investigation of the damaged zone created by perforating

    International Nuclear Information System (INIS)

    Pucknell, J.K.; Behrmann, L.A.

    1991-01-01

    This paper reports on underbalance perforation flow experiments performed on reservoir and outcrop sandstones to investigate the perforation damaged zone. Cores from several different formations were perforated under reservoir conditions. After perforating, the cores were examined using CAT scans (Computer Aided tomography), thin sections and mercury porosimetry. In conjunction with these measurements, permeabilities in the damaged zone were measured using a minipermeameter and radial flow permeameter or were estimated from pore size distribution. The density and porosity of the damaged zone (at least for saturated rocks) is essentially the same as that in the undamaged rock. The damaged zone is not compacted, contrary to suggestions made in earlier work. However, the creation of this zone involves the destruction of large pores. The volume lost from these pores is replaced by microfractures created when rock grains are fractured by penetration of the shaped charge jet. This reduction in the average pore size causes a reduction in the permeability with the damaged zone. Although direct measurement of this permeability was made difficult by naturally occurring permeability variations, unambiguous measurements were obtained

  12. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1. Volume 5: Analysis of core damage frequency from seismic events during mid-loop operations

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

    1994-08-01

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1) and the other at Sandia National Laboratories studying a boiling water reactor (Grand Gulf). Both the Brookhaven and Sandia projects have examined only accidents initiated by internal plant faults--so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Surry Unit 1. All of the many systems modeling assumptions, component non-seismic failure rates, and human error rates that were used in the internal-initiator study at Surry have been adopted here, so that the results of the two studies can be as comparable as possible. Both the Brookhaven study and this study examine only two shutdown plant operating states (POSs) during refueling outages at Surry, called POS 6 and POS 10, which represent mid-loop operation before and after refueling, respectively. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POSs 6 and 10. The results of the analysis are that the core-damage frequency of earthquake-initiated accidents during refueling outages in POS 6 and POS 10 is found to be low in absolute terms, less than 10 -6 /year

  13. Effect of fluid-to-structure heat transfer on the structural damage potential to a liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hakim, S.J.; Abramson, P.B.

    1979-01-01

    Deterministic calculations simulating a hypothetical accident in a liquid-metal fast breeder reactor that leads to a hydrodynamic disassembly of the core have been carried out to estimate the system's damage potential due to the vapor-pressure-driven expansion of molten core material and its dependency on the heat transfer to the remaining structure. These calculations ignored the effect on the work potential of sodium left in the core during the disassembly. Results indicate that steel cladding in the upper axial blankets and fission gas plenum acts as a thermodynamic energy sink that could reduce the total thermodynamic work energy by between one and two orders of magnitude, provided little or no sodium is present in the core at the time of interaction. These results have been found to be insensitive to the rate of heat transferred from the molten fuel to the molten steel that comprises the molten core material

  14. Stiffness and strength degradation of damaged truss core composites

    Czech Academy of Sciences Publication Activity Database

    Šiška, Filip; Tawfeeq, Arwa F.; Dlouhý, I.; Barnett, M.R.

    2015-01-01

    Roč. 125, JUL (2015), s. 287-294 ISSN 0263-8223 R&D Projects: GA MŠk EE2.3.20.0197 Institutional support: RVO:68081723 Keywords : Truss core composites * Finite element * Strain rate * High temperature tests Subject RIV: JI - Composite Materials Impact factor: 3.853, year: 2015

  15. TMI-2 core examination

    International Nuclear Information System (INIS)

    Hobbins, R.R.; MacDonald, P.E.; Owen, D.E.

    1983-01-01

    The examination of the damaged core at the Three Mile Island Unit 2 (TMI-2) reactor is structured to address the following safety issues: fission product release, transport, and deposition; core coolability; containment integrity; and recriticality during severe accidents; as well as zircaloy cladding ballooning and oxidation during so-called design basis accidents. The numbers of TMI-2 components or samples to be examined, the priority of each examination, the safety issue addressed by each examination, the principal examination techniques to be employed, and the data to be obtained and the principal uses of the data are discussed in this paper

  16. High Power Spark Delivery System Using Hollow Core Kagome Lattice Fibers

    Directory of Open Access Journals (Sweden)

    Ciprian Dumitrache

    2014-08-01

    Full Text Available This study examines the use of the recently developed hollow core kagome lattice fibers for delivery of high power laser pulses. Compared to other photonic crystal fibers (PCFs, the hollow core kagome fibers have larger core diameter (~50 µm, which allows for higher energy coupling in the fiber while also maintaining high beam quality at the output (M2 = 1.25. We have conducted a study of the maximum deliverable energy versus laser pulse duration using a Nd:YAG laser at 1064 nm. Pulse energies as high as 30 mJ were transmitted for 30 ns pulse durations. This represents, to our knowledge; the highest laser pulse energy delivered using PCFs. Two fiber damage mechanisms were identified as damage at the fiber input and damage within the bulk of the fiber. Finally, we have demonstrated fiber delivered laser ignition on a single-cylinder gasoline direct injection engine.

  17. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Analysis of core damage frequency from internally induced flooding events for Plant Operational State 5 during a refueling outage. Volume 4

    International Nuclear Information System (INIS)

    Dandini, V.; Staple, B.; Kirk, H.; Whitehead, D.; Forester, J.

    1994-07-01

    An estimate of the contribution of internal flooding to the mean core damage frequency at the Grand Gulf Nuclear Station was calculated for Plant Operational State 5 during a refueling outage. Pursuant to this objective, flood zones and sources were identified and flood volumes were calculated. Equipment necessary for the maintenance of plant safety was identified and its vulnerability to flooding was determined. Event trees and fault trees were modified or developed as required, and PRA quantification was performed using the IRRAS code. The mean core damage frequency estimate for GGNS during POS 5 was found to be 2.3 E-8 per year

  18. An improved method for quantitatively measuring the sequences of total organic carbon and black carbon in marine sediment cores

    Science.gov (United States)

    Xu, Xiaoming; Zhu, Qing; Zhou, Qianzhi; Liu, Jinzhong; Yuan, Jianping; Wang, Jianghai

    2018-01-01

    Understanding global carbon cycle is critical to uncover the mechanisms of global warming and remediate its adverse effects on human activities. Organic carbon in marine sediments is an indispensable part of the global carbon reservoir in global carbon cycling. Evaluating such a reservoir calls for quantitative studies of marine carbon burial, which closely depend on quantifying total organic carbon and black carbon in marine sediment cores and subsequently on obtaining their high-resolution temporal sequences. However, the conventional methods for detecting the contents of total organic carbon or black carbon cannot resolve the following specific difficulties, i.e., (1) a very limited amount of each subsample versus the diverse analytical items, (2) a low and fluctuating recovery rate of total organic carbon or black carbon versus the reproducibility of carbon data, and (3) a large number of subsamples versus the rapid batch measurements. In this work, (i) adopting the customized disposable ceramic crucibles with the microporecontrolled ability, (ii) developing self-made or customized facilities for the procedures of acidification and chemothermal oxidization, and (iii) optimizing procedures and carbon-sulfur analyzer, we have built a novel Wang-Xu-Yuan method (the WXY method) for measuring the contents of total organic carbon or black carbon in marine sediment cores, which includes the procedures of pretreatment, weighing, acidification, chemothermal oxidation and quantification; and can fully meet the requirements of establishing their highresolution temporal sequences, whatever in the recovery, experimental efficiency, accuracy and reliability of the measurements, and homogeneity of samples. In particular, the usage of disposable ceramic crucibles leads to evidently simplify the experimental scenario, which further results in the very high recovery rates for total organic carbon and black carbon. This new technique may provide a significant support for

  19. Materials behaviour in PWRs core

    International Nuclear Information System (INIS)

    Barbu, A.; Massoud, J.P.

    2008-01-01

    Like in any industrial facility, the materials of PWR reactors are submitted to mechanical, thermal or chemical stresses during particularly long durations of operation: 40 years, and even 60 years. Materials closer to the nuclear fuel are submitted to intense bombardment of particles (mainly neutrons) coming from the nuclear reactions inside the core. In such conditions, the damages can be numerous and various: irradiation aging, thermal aging, friction wear, generalized corrosion, stress corrosion etc.. The understanding of the materials behaviour inside the cores of reactors in operation is a major concern for the nuclear industry and its long term forecast is a necessity. This article describes the main ways of materials degradation without and under irradiation, with the means used to foresee their behaviour using physics-based models. Content: 1 - structures, components and materials: structure materials, nuclear materials; 2 - main ways of degradation without irradiation: thermal aging, stress corrosion, wear; 3 - main ways of degradation under irradiation: microscopic damaging - point defects, dimensional alterations, evolution of mechanical characteristics under irradiation, irradiation-assisted stress corrosion cracking (IASCC), synergies; 4 - forecast of materials evolution under irradiation using physics-based models: primary damage - fast dynamics, primary damage annealing - slow kinetics microstructural evolution, impact of microstructural changes on the macroscopic behaviour, insight on modeling methods; 5 - materials change characterization techniques: microscopic techniques - direct defects observation, nuclear techniques using a particle beam, global measurements, mechanical characterizations; 6 - perspectives. (J.S.)

  20. Risk of nuclear damage

    International Nuclear Information System (INIS)

    Kienzl, K.

    1997-01-01

    Following the opening and words of welcome by Mr. Fritz Unterpertinger (unit director at the Austrian Federal Ministry for the Environment, Youth and Family; BMUJF) Mrs Helga Kromp-Kolb (professor at the Institute for Meteorology and Physics of the University of Natural Resources Science Vienna) illustrated the risks of nuclear damage in Europe by means of a nuclear risk map. She explained that even from a scientific or technical point of view the assessment of risks arising from nuclear power stations was fraught with great uncertainties. Estimates about in how far MCAs (maximum credible accident) could still be controlled by safety systems vary widely and so do assessments of the probability of a core melt. But there is wide agreement in all risk assessments conducted so far that MCAs might occur within a - from a human point of view - conceivable number of years. In this connection one has to bear in mind that the occurrence of such a major accident - whatever its probability may be - could entail immense damage and the question arises whether or not it is at all justifiable to expose the general public to such a risk. Klaus Rennings (Centre for European Economic Research, Mannheim, Germany) dealt with the economic aspects of nuclear risk assessment. He explained that there are already a number of studies available aiming to assess the risk of damage resulting from a core melt accident in economic terms. As to the probability of occurrence estimates vary widely between one incident in 3,333 and 250,000 year of reactor operation. It is assumed, however, that a nuclear accident involving a core melt in Germany would probably exceed the damage caused by the Chernobyl accident. The following speakers addressed the legal aspects of risks associated with nuclear installations. Mrs Monika Gimpel-Hinteregger (professor at the Institute for Civil Law in Graz) gave an overview on the applicable Austrian law concerning third party liability in the field of nuclear energy

  1. Method for orienting a borehole core

    International Nuclear Information System (INIS)

    Henry, W.

    1980-01-01

    A method is described for longitudinally orienting a borehold core with respect to the longitudinal axis of the drill string which drilled said borehold core in such a manner that the original longitudinal attitude of said borehold core within the earth may be determined. At least a portion of said borehold core is partialy demagnetized in steps to thereby at least partially remove in steps the artificial remanent magnetism imparted to said borehole core by said drill string. The artifical remanent magnetism is oriented substantially parallel to the longitudinal axis of said drill string. The direction and intensity of the total magnetism of said borehold core is measured at desired intervals during the partial demagnetizing procedure. An artificial remanent magnetism vector is established which extends from the final measurement of the direction and intensity of the total magnetism of said borehole core taken during said partial demagnetizing procedure towards the initial measurement of the direction and intensity of the total magnetism of said borehold core taken during said partial demagnetizing procedure. The borehold core is oriented in such a manner that said artificial remanent magnetism vector points at least substantially downwardly towards the bottom of said borehold core for a borehold in the northern hemisphere and points at least substantailly upwardly towards the top of said borehole core for a borehole in the southern hemisphere

  2. TMI-2 core boring machine

    International Nuclear Information System (INIS)

    Croft, K.M.; Helbert, H.J.; Laney, W.M.

    1986-01-01

    An important and essential aspect of the TMI-2 defueling effort is to determine what occurred in the core region during the accident. Remote cameras and probes only portray a portion of the overall picture. What lies beneath the rubble bed and solidified sublayer is, as yet, unknown. This paper discusses the TMI-2 Core Boring Machine, which has been developed to drill into the damaged core of the TMI-2 reactor and extract stratified samples of the core. This machine, its unique support structure, positioning and leveling systems, and specially designed drill bits, combine to provide a unique mechanical system. In addition, the machine is controlled by a microprocessor; which actually controls the drilling operation, allowing relatively inexperienced operators to drill the core samples. A data acquisition system is data integral with the controlling system and collects data relative to system conditions and monitored parameters during drilling. Data obtained during the actual drilling operations are collected in a data base which will be used for actual mapping of the core region, identifying materials and stratification levels that are present

  3. Core baffle for nuclear reactors

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1977-01-01

    The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

  4. Calculation of Core Damage Frequency for the Change of the Common Cause Failure Parameters According to the Testing Strategies

    International Nuclear Information System (INIS)

    Kang, Dae Il; Kim, Kil You; Jin, Young Ho; Kim, Tae Woon

    2011-01-01

    Common cause failure (CCF) probabilities are differently estimated according to testing strategies. There are two representative testing schemes; staggered testing and non-staggered testing schemes. For the cases where trains or channels of standby safety systems consisting of more than two redundant components are tested in a staggered manner, the standby safety components within a train can be tested simultaneously or consecutively. In this case, mixed testing scheme, staggered and non-staggered testing schemes, are used for testing the components. Kang et al. derived the formulas for the estimations of the CCF probabilities of the components under the mixed testing scheme. This paper presents the sensitivity study results on the core damage frequency (CDF) of the SMART (System-integrated Modular Advanced Reactor) for the changes of the CCF parameters according to the testing strategies

  5. Using CdTe/ZnSe core/shell quantum dots to detect DNA and damage to DNA

    Directory of Open Access Journals (Sweden)

    Moulick A

    2017-02-01

    Full Text Available Amitava Moulick,1,2 Vedran Milosavljevic,1,2 Jana Vlachova,1,2 Robert Podgajny,3 David Hynek,1,2 Pavel Kopel,1,2 Vojtech Adam1,2 1Department of Chemistry and Biochemistry, Mendel University, 2Central European Institute of Technology, Brno University of Technology, Brno, Czech Republic; 3Faculty of Chemistry, Jagiellonian University, Krakow, Poland Abstract: CdTe/ZnSe core/shell quantum dot (QD, one of the strongest and most highly luminescent nanoparticles, was directly synthesized in an aqueous medium to study its individual interactions with important nucleobases (adenine, guanine, cytosine, and thymine in detail. The results obtained from the optical analyses indicated that the interactions of the QDs with different nucleobases were different, which reflected in different fluorescent emission maxima and intensities. The difference in the interaction was found due to the different chemical behavior and different sizes of the formed nanoconjugates. An electrochemical study also confirmed that the purines and pyrimidines show different interactions with the core/shell QDs. Based on these phenomena, a novel QD-based method is developed to detect the presence of the DNA, damage to DNA, and mutation. The QDs were successfully applied very easily to detect any change in the sequence (mutation of DNA. The QDs also showed their ability to detect DNAs directly from the extracts of human cancer (PC3 and normal (PNT1A cells (detection limit of 500 pM of DNA, which indicates the possibilities to use this easy assay technique to confirm the presence of living organisms in extreme environments. Keywords: nanoparticles, nucleobases, biosensor, fluorescence, mutation

  6. Estimative of core damage frequency in IPEN's IEA-R1 research reactor (PSA level 1) due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami

    2009-01-01

    This work applies the methodology of probabilistic safety assessment level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid by major pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, emergency core cooling system (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  7. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  8. Windscale pile core surveys

    International Nuclear Information System (INIS)

    Curtis, R.F.; Mathews, R.F.

    1996-01-01

    The two Windscale Piles were closed down, defueled as far as possible and mothballed for thirty years following a fire in the core of Pile 1 in 1957 resulting from the spontaneous release of stored Wigner energy in the graphite moderator. Decommissioning of the reactors commenced in 1987 and has reached the stage where the condition of both cores needs to be determined. To this end, non-intrusive and intrusive surveys and sampling of the cores have been planned and partly implemented. The objectives for each Pile differ slightly. The location and quantity of fuel remaining in the damaged core of Pile 1 needed to be established, whereas the removal of all fuel from Pile 2 needed to be confirmed. In Pile 1, the possible existence of a void in the core is to be explored and in Pile 2, the level of Wigner energy remaining required to be quantified. Levels of radioactivity in both cores needed to be measured. The planning of the surveys is described including strategy, design, safety case preparation and the remote handling and viewing equipment required to carry out the inspection, sampling and monitoring work. The results from the completed non-intrusive survey of Pile 2 are summarised. They confirm that the core is empty and the graphite is in good condition. The survey of Pile 1 has just started. (UK)

  9. Review of Repair Materials for Fire-Damaged Reinforced Concrete Structures

    Science.gov (United States)

    Zahid, MZA Mohd; Abu Bakar, BH; Nazri, FM; Ahmad, MM; Muhamad, K.

    2018-03-01

    Reinforced concrete (RC) structures perform well during fire and may be repaired after the fire incident because their low heat conductivity prevents the loss or degradation of mechanical strength of the concrete core and internal reinforcing steel. When an RC structure is heated to more than 500 °C, mechanical properties such as compressive strength, stiffness, and tensile strength start to degrade and deformations occur. Although the fire-exposed RC structure shows no visible damage, its residual strength decreases compared with that in the pre-fire state. Upon thorough assessment, the fire-damaged RC structure can be repaired or strengthened, instead of subjecting to partial or total demolition followed by reconstruction. The structure can be repaired using several materials, such as carbon fiber-reinforced polymer, glass fiber-reinforced polymer, normal strength concrete, fiber-reinforced concrete, ferrocement, epoxy resin mortar, and high-performance concrete. Selecting an appropriate repair material that must be compatible with the substrate or base material is a vital step to ensure successful repair. This paper reviews existing repair materials and factors affecting their performance. Of the materials considered, ultra-high-performance fiber-reinforced concrete (UHPFRC) exhibits huge potential for repairing fire-damaged RC structures but lack of information available. Hence, further studies must be performed to assess the potential of UHPFRC in rehabilitating fire-damaged RC structures.

  10. The assessment of the integrity of AGR core during an earthquake

    International Nuclear Information System (INIS)

    Smith, C.R.

    1987-01-01

    The seismic response of the core has been calculated using an idealisation having several hundred thousand degrees of freedom. The individual graphite bricks are idealised as rigid masses, whilst contact spring elements are used to represent the load transmissions or impacts that can take place between the bricks. The necessary input information for the contact spring elements (i.e. stiffness, damping and friction), has been obtained from test work. Whilst the dynamic response of the core itself is non-linear, the supporting steel structures are linearly elastic. Consequently, the dynamic characteristics of the supporting structures are evaluated with the non-linear core structure uncoupled, and are then used with the non-linear core model in a step-by-step explicit time history analysis. The paper discusses the analytical model and presents results from some of the predictions of core dynamic response to earthquakes. The development of criteria for graphite impacts, based on the J integral, is described. Impact tests on a range of brick slices have been used to give data on brick or key cracking under repeated impacts. Dynamic analysis of plane stress finite element models of these test geometries has been carried out in order to establish a qualified analysis method which can be used to extrapolate the test data to impact damage in the core. This analysis method is applied to finite element models of the core bricks in which the loadings due to operating conditions, environmental and ageing effects are included. In the presence of any existing state of stress at any time during the operating life, the damage due to repeated impacts defined by the time-history seismic response of the core may then be estimated through a cumulative damage procedure. (author)

  11. Controlled fabrication of multi-core alginate microcapsules.

    Science.gov (United States)

    Eqbal, Md Danish; Gundabala, Venkat

    2017-12-01

    In this work, we present a robust microfluidic platform for controlled and complete on-chip generation of alginate microcapsules with single and double liquid cores. A combined Coflow and T-junction configuration implemented in a hybrid glass-PDMS (Polydimethylsiloxane) device is used for the generation of microcapsules with oil as liquid core. Frequency matching of oil-alginate double emulsion generation with that of aqueous Calcium chloride droplet generation allows for controlled merging of the two, resulting in reliable production of microcapsules. Confocal imaging of microcapsule cross-section reveals presence of intact liquid core. In the case of double core microcapsules, the two cores are well separated by alginate layer ensuring their long term stability. The current approach is expected to have advantages over existing techniques for liquid core microcapsule generation in terms of continuity of the process, control over core stability, and non-damage to cells when used for cell encapsulation applications. Copyright © 2017 Elsevier Inc. All rights reserved.

  12. The Damage and Geochemical Signature of a Crustal Scale Strike-Slip Fault Zone

    Science.gov (United States)

    Gomila, R.; Mitchell, T. M.; Arancibia, G.; Jensen Siles, E.; Rempe, M.; Cembrano, J. M.; Faulkner, D. R.

    2013-12-01

    Fluid-flow migration in the upper crust is strongly controlled by fracture network permeability and connectivity within fault zones, which can lead to fluid-rock chemical interaction represented as mineral precipitation in mesh veins and/or mineralogical changes (alteration) of the host rock. While the dimensions of fault damage zones defined by fracture intensity is beginning to be better understood, how such dimensions compare to the size of alteration zones is less well known. Here, we show quantitative structural and chemical analyses as a function of distance from a crustal-scale strike-slip fault in the Atacama Fault System, Northern Chile, to compare fault damage zone characteristics with its geochemical signature. The Jorgillo Fault (JF) is a ca. 18 km long NNW striking strike-slip fault cutting Mesozoic rocks with sinistral displacement of ca. 4 km. In the study area, the JF cuts through orthogranulitic and gabbroic rocks at the west (JFW) and the east side (JFE), respectively. A 200 m fault perpendicular transect was mapped and sampled for structural and XRF analyses of the core, damage zone and protolith. The core zone consists of a ca. 1 m wide cataclasite zone bounded by two fault gouge zones ca. 40 cm. The damage zone width defined by fracture density is ca. 50 m wide each side of the core. The damage zone in JFW is characterized by NW-striking subvertical 2 cm wide cataclastic rocks and NE-striking milimetric open fractures. In JFE, 1-20 mm wide chlorite, quartz-epidote and quartz-calcite veins, cut the gabbro. Microfracture analysis in JFW reveal mm-wide cataclasitic/ultracataclasitic bands with clasts of protolith and chlorite orientated subparallel to the JF in the matrix, calcite veins in a T-fractures orientation, and minor polidirectional chlorite veins. In JFE, chlorite filled conjugate fractures with syntaxial growth textures and evidence for dilational fracturing processes are seen. Closest to the core, calcite veins crosscut chlorite veins

  13. [Core muscle chains activation during core exercises determined by EMG-a systematic review].

    Science.gov (United States)

    Rogan, Slavko; Riesen, Jan; Taeymans, Jan

    2014-10-15

    Good core muscles strength is essential for daily life and sports activities. However, the mechanism how core muscles may be effectively triggered by exercises is not yet precisely described in the literature. The aim of this systematic review was to evaluate the rate of activation as measured by electromyography of the ventral, lateral and dorsal core muscle chains during core (trunk) muscle exercises. A total of 16 studies were included. Exercises with a vertical starting position, such as the deadlift or squat activated significantly more core muscles than exercises in the horizontal initial position.

  14. In Vitro-Assembled Alphavirus Core-Like Particles Maintain a Structure Similar to That of Nucleocapsid Cores in Mature Virus

    OpenAIRE

    Mukhopadhyay, Suchetana; Chipman, Paul R.; Hong, Eunmee M.; Kuhn, Richard J.; Rossmann, Michael G.

    2002-01-01

    In vitro-assembled core-like particles produced from alphavirus capsid protein and nucleic acid were studied by cryoelectron microscopy. These particles were found to have a diameter of 420 Å with 240 copies of the capsid protein arranged in a T=4 icosahedral surface lattice, similar to the nucleocapsid core in mature virions. However, when the particles were subjected to gentle purification procedures, they were damaged, preventing generation of reliable structural information. Similarly, pu...

  15. Compendium of Current Total Ionizing Dose and Displacement Damage Results from NASA Goddard Space Flight Center and Selected NASA Electronic Parts and Packaging Program

    Science.gov (United States)

    Topper, Alyson D.; Campola, Michael J.; Chen, Dakai; Casey, Megan C.; Yau, Ka-Yen; Cochran, Donna J.; LaBel, Kenneth A.; Ladbury, Raymond L.; Lauenstein, Jean-Marie; Mondy, Timothy K.; hide

    2017-01-01

    Total ionizing dose and displacement damage testing was performed to characterize and determine the suitability of candidate electronics for NASA space utilization. Devices tested include optoelectronics, digital, analog, linear bipolar devices, and hybrid devices.

  16. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  17. Estimative of core damage frequency in IPEN'S IEA-R1 research reactor due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami; Sabundjian, Gaiane; Cabral, Eduardo Lobo Lustosa

    2009-01-01

    The National Commission of Nuclear Energy (CNEN), which is the Brazilian nuclear regulatory commission, imposes safety and licensing standards in order to ensure that the nuclear power plants operate in a safe way. For licensing a nuclear reactor one of the demands of CNEN is the simulation of some accidents and thermalhydraulic transients considered as design base to verify the integrity of the plant when submitted to adverse conditions. The accidents that must be simulated are those that present large probability to occur or those that can cause more serious consequences. According to the FSAR (Final Safety Analysis Report) the initiating event that can cause the largest damage in the core, of the IEA-R1 research reactor at IPEN-CNEN/SP, is the LOCA (Loss of Coolant Accident). The objective of this paper is estimate the frequency of the IEA-R1 core damage, caused by this initiating event. In this paper we analyze the accident evolution and performance of the systems which should mitigate this event: the Emergency Coolant Core System (ECCS) and the isolated pool system. They will be analyzed by means of the event tree. In this work the reliability of these systems are also quantified using the fault tree. (author)

  18. Drilling history core hole DC-8

    International Nuclear Information System (INIS)

    1978-10-01

    Core hole DC-8 was completed in August, 1978 by Boyles Brothers Drilling Company, Spokane, Washington, under subcontract to Fenix and Scission, Inc. The hole was cored for the US Department of Energy and the Rockwell Hanford Operations' Basalt Waste Isolation Program. Fenix and Scisson, Inc. furnished the engineering, daily supervision of the core drilling activities, and geologic core logging for hole DC-8. Core hole DC-8 is located on the Hanford Site near the Wye Barricade and 50 feet northwest of rotary hole DC-7. The Hanford Site vation coordinates for DC-8 are North 14,955.94 feet and West 14,861.92 coordinates for DC-8 are North 14,955.94 feet and West 14,861.92 mean sea level. The purpose of core hole DC-8 was to core drill vertically through the basalt and interbed units for stratigraphic depth determination and core collection, and to provide a borehole for hydrologic testing and cross-hole seismic shear and pressure wave velocity studies with rotary hole DC-7. The total depth of core hole DC-8 was 4100.5 feet. Core recovery exceeded 97 percent of the total footage cored

  19. Drilling history core hole DC-8

    Energy Technology Data Exchange (ETDEWEB)

    1978-10-01

    Core hole DC-8 was completed in August, 1978 by Boyles Brothers Drilling Company, Spokane, Washington, under subcontract to Fenix and Scission, Inc. The hole was cored for the US Department of Energy and the Rockwell Hanford Operations' Basalt Waste Isolation Program. Fenix and Scisson, Inc. furnished the engineering, daily supervision of the core drilling activities, and geologic core logging for hole DC-8. Core hole DC-8 is located on the Hanford Site near the Wye Barricade and 50 feet northwest of rotary hole DC-7. The Hanford Site vation coordinates for DC-8 are North 14,955.94 feet and West 14,861.92 coordinates for DC-8 are North 14,955.94 feet and West 14,861.92 mean sea level. The purpose of core hole DC-8 was to core drill vertically through the basalt and interbed units for stratigraphic depth determination and core collection, and to provide a borehole for hydrologic testing and cross-hole seismic shear and pressure wave velocity studies with rotary hole DC-7. The total depth of core hole DC-8 was 4100.5 feet. Core recovery exceeded 97 percent of the total footage cored.

  20. ROSA full-core and DNBR capabilities

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2013-01-01

    The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  1. ROSA full-core and DNBR capabilities

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2012-01-01

    This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  2. Numerical study on core damage and interpretation of in situ state of stress

    Energy Technology Data Exchange (ETDEWEB)

    Hakala, M. [Gridpoint Finland Oy (Finland)

    1999-06-01

    Core disking is a phenomenon where a diamond cored core sample will be sliced when released from a stressed host rock. Ring disking is a similar phenomenon which takes place during overcoring with a pilot hole. Because of the uniform shape and spacing of disk fracturing, it has the potential to be used for estimating the in situ state of stress. If this is feasible, it could be used in high stress states where the traditional stress measuring techniques are not valid or even possible. In this work the both the core disking and ring disking phenomena were studied based on the elastic bottom hole stress application developed and a series of fracture growth stability simulations. The results-showed that both phenomena are very complicated and site specific, but the spacing, shape, extent and initiation point are clearly stress state dependent. Throughout the work, guidelines for the in situ stress field interpretation method were developed and implemented for the borehole aligned orthogonal stress field and Poisson`s ratio of 0.25. Based on this study, the in situ state of stress can be estimated with acceptable accuracy if information on both core disking and ring disking is available. On the other hand, as an indirect method, there are no reasons to use it if direct measurements can be used. (orig.) 35 refs.

  3. Numerical study on core damage and interpretation of in situ state of stress

    International Nuclear Information System (INIS)

    Hakala, M.

    1999-06-01

    Core disking is a phenomenon where a diamond cored core sample will be sliced when released from a stressed host rock. Ring disking is a similar phenomenon which takes place during overcoring with a pilot hole. Because of the uniform shape and spacing of disk fracturing, it has the potential to be used for estimating the in situ state of stress. If this is feasible, it could be used in high stress states where the traditional stress measuring techniques are not valid or even possible. In this work the both the core disking and ring disking phenomena were studied based on the elastic bottom hole stress application developed and a series of fracture growth stability simulations. The results-showed that both phenomena are very complicated and site specific, but the spacing, shape, extent and initiation point are clearly stress state dependent. Throughout the work, guidelines for the in situ stress field interpretation method were developed and implemented for the borehole aligned orthogonal stress field and Poisson's ratio of 0.25. Based on this study, the in situ state of stress can be estimated with acceptable accuracy if information on both core disking and ring disking is available. On the other hand, as an indirect method, there are no reasons to use it if direct measurements can be used. (orig.)

  4. Development of Structural Core Components for Breeder Reactors

    International Nuclear Information System (INIS)

    Saibaba, N.

    2013-01-01

    Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

  5. Impact damage in aircraft composite sandwich panels

    Science.gov (United States)

    Mordasky, Matthew D.

    An experimental study was conducted to develop an improved understanding of the damage caused by runway debris and environmental threats on aircraft structures. The velocities of impacts for stationary aircraft and aircraft under landing and takeoff speeds was investigated. The impact damage by concrete, asphalt, aluminum, hail and rubber sphere projectiles was explored in detail. Additionally, a kinetic energy and momentum experimental study was performed to look at the nature of the impacts in more detail. A method for recording the contact force history of the impact by an instrumented projectile was developed and tested. The sandwich composite investigated was an IM7-8552 unidirectional prepreg adhered to a NOMEXRTM core with an FM300K film adhesive. Impact experiments were conducted with a gas gun built in-house specifically for delivering projectiles to a sandwich composite target in this specic velocity regime (10--140 m/s). The effect on the impact damage by the projectile was investigated by ultrasonic C-scan, high speed camera and scanning electron and optical microscopy. Ultrasonic C-scans revealed the full extent of damage caused by each projectile, while the high speed camera enabled precise projectile velocity measurements that were used for striking velocity, kinetic energy and momentum analyses. Scanning electron and optical images revealed specific features of the panel failure and manufacturing artifacts within the lamina and honeycomb core. The damage of the panels by different projectiles was found to have a similar damage area for equivalent energy levels, except for rubber which had a damage area that increased greatly with striking velocity. Further investigation was taken by kinetic energy and momentum based comparisons of 19 mm diameter stainless steel sphere projectiles in order to examine the dominating damage mechanisms. The sandwich targets were struck by acrylic, aluminum, alumina, stainless steel and tungsten carbide spheres of the

  6. Flexural fatigue failures and lives of Eco-Core sandwich beams

    International Nuclear Information System (INIS)

    Hossain, Mohammad Mynul; Shivakumar, Kunigal

    2014-01-01

    Highlights: • Eco-Core sandwich beam is flexural fatigue tested to study its fatigue response. • The core showed three failure types: damage onset, progression and final failure. • These failures were found to be represented by 1%, 5% and 7% change in compliance. • The fatigue stress-life (S–N) relationship follows a power low, σ max /σ ct = A o N α . • The fatigue failure was by multiple vertical cracks followed by 45° shear failure. - Abstract: Eco-Core is a class of syntactic foam made from small volume of high char yield binder and large volume of a class of flyash for fire resistance application. Very little or no flexural fatigue data of this class of core material is reported in the open literature. This paper presents a flexural fatigue response of Eco-Core in a glass/vinyl ester composite face sheet sandwich beam. A four-point loaded flexural test specimen was designed and tested in static and fatigue loadings to cause tension failure in the core. The fatigue test was conducted at maximum cyclic stress (σ max ) ranged from 0.7σ ct to 0.9σ ct , where σ ct is the static flexural strength of the core. The sinusoidal loading frequency of 2 Hz with the stress ratio of 0.1 was used. Flexural fatigue failure modes of Eco-Core sandwich beam were classified: damage onset (single tension crack), damage progression (multiple tension cracks) and ultimate failure (a combination of tension and shear). These failures were characterized by 1%, 5% and 7% changes in compliance that corresponds to N 1% , N 5% and N 7% lives. The fatigue stress-life (S–N) relationship was found to follow the well-known power law equation, σ max /σ ct = A o N α . The constants A o and α were established for all three types of failures. The endurance limit was established based on 1 million cycles limit and it was found to be 0.65σ ct , 0.70σ ct and 0.71σ ct , respectively for the three modes of failure. Flexural fatigue and static failure modes of Eco-Core sandwich

  7. The potential of permeability damage during thermal recovery of Cold Lake bitumen

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Wiwchar, B.; Gunter, W. D. [Alberta Research Council, Devon, AB (Canada); Dudley, J. S. [Imperial Oil Resources, Calgary, AB (Canada)

    1999-09-01

    Methods and results of coreflood tests designed to evaluate permeability damage caused by Clearwater formation clays in the Cold Lake area of Alberta are described. Three periods of permeability damage were encountered, the first during and shortly after the core was heated to 250 degrees C. Experimental evidence suggests that thermally activated grain crushing and subsequent fines migration were responsible for this initial permeability loss. The second period of damage was a gradual process which resulted in 65 per cent and 78 percent of permeability loss for the two corefloods, respectively. This phase of the permeability damage was considered to have been the result of hydrothermal reactions (berthierine to Fe-saponite). The third period of permeability damage occurred when fresh water was injected into the core. This was attributed to osmotic swelling of the Fe-saponite. A comparison of field evidence with experimental results revealed certain discrepancies, suspected to be due to the kinetics of the reaction, including disruption of berthierine grain coats and permeability damage due to subsequent fines migration. To err on the safe side, it is recommended that thermal recovery wells should be completed away from berthierine-rich zones. 15 refs., 2 tabs., 7 figs.

  8. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  9. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Analysis of core damage frequency from internal fire events for Plant Operational State 5 during a refueling outage. Volume 3

    International Nuclear Information System (INIS)

    Lambright, J.; Yakle, J.

    1994-07-01

    This report, Volume 3, presents the details of the analysis of core damage frequency due to fire during shutdown Plant Operational State 5 at the Grand Gulf Nuclear Station. Insights from previous fire analyses (Peach Bottom, Surry, LaSalle) were used to the greatest extent possible in this analysis. The fire analysis was fully integrated utilizing the same event trees and fault trees that were used in the internal events analysis. In assessing shutdown risk due to fire at Grand Gulf, a detailed screening was performed which included the following elements: (a) Computer-aided vital area analysis; (b) Plant inspections; (c) Credit for automatic fire protection systems; (d) Recovery of random failures; (e) Detailed fire propagation modeling. This screening process revealed that all plant areas had a negligible (<1.0E-8 per year) contribution to fire-induced core damage frequency

  10. Severe fuel damage investigations of KFK/PNS

    International Nuclear Information System (INIS)

    Fiege, A.

    1983-01-01

    This report is a comprehensive review of the objectives, the program planning, the status and the further procedure of the investigations of KfK/PNS on severe core damage. The investigations were started in 1981 and will be finished in 1985/86. (orig.) [de

  11. Total intermittent Pringle maneuver during liver resection can induce intestinal epithelial cell damage and endotoxemia.

    Directory of Open Access Journals (Sweden)

    Simon A W G Dello

    Full Text Available OBJECTIVES: The intermittent Pringle maneuver (IPM is frequently applied to minimize blood loss during liver transection. Clamping the hepatoduodenal ligament blocks the hepatic inflow, which leads to a non circulating (hepatosplanchnic outflow. Also, IPM blocks the mesenteric venous drainage (as well as the splenic drainage with raising pressure in the microvascular network of the intestinal structures. It is unknown whether the IPM is harmful to the gut. The aim was to investigate intestinal epithelial cell damage reflected by circulating intestinal fatty acid binding protein levels (I-FABP in patients undergoing liver resection with IPM. METHODS: Patients who underwent liver surgery received total IPM (total-IPM or selective IPM (sel-IPM. A selective IPM was performed by selectively clamping the right portal pedicle. Patients without IPM served as controls (no-IPM. Arterial blood samples were taken immediately after incision, ischemia and reperfusion of the liver, transection, 8 hours after start of surgery and on the first post-operative day. RESULTS: 24 patients (13 males were included. 7 patients received cycles of 15 minutes and 5 patients received cycles of 30 minutes of hepatic inflow occlusion. 6 patients received cycles of 15 minutes selective hepatic occlusion and 6 patients underwent surgery without inflow occlusion. Application of total-IPM resulted in a significant increase in I-FABP 8 hours after start of surgery compared to baseline (p<0.005. In the no-IPM group and sel-IPM group no significant increase in I-FABP at any time point compared to baseline was observed. CONCLUSION: Total-IPM in patients undergoing liver resection is associated with a substantial increase in arterial I-FABP, pointing to intestinal epithelial injury during liver surgery. TRIAL REGISTRATION: ClinicalTrials.gov NCT01099475.

  12. Adolescents' non-core food intake: a description of what, where and with whom adolescents consume non-core foods.

    Science.gov (United States)

    Toumpakari, Zoi; Haase, Anne M; Johnson, Laura

    2016-06-01

    Little is known about adolescents' non-core food intake in the UK and the eating context in which they consume non-core foods. The present study aimed to describe types of non-core foods consumed by British adolescents in total and across different eating contexts. A descriptive analysis, using cross-sectional data from food diaries. Non-core foods were classified based on cut-off points of fat and sugar from the Australian Guide to Healthy Eating. Eating context was defined as 'where' and 'with whom' adolescents consumed each food. Percentages of non-core energy were calculated for each food group in total and across eating contexts. A combined ranking was then created to account for each food's contribution to non-core energy intake and its popularity of consumption (percentage of consumers). The UK National Diet and Nutrition Survey 2008-2011. Adolescents across the UK aged 11-18 years (n 666). Non-core food comprised 39·5 % of total energy intake and was mostly 'Regular soft drinks', 'Crisps & savoury snacks', 'Chips & potato products', 'Chocolate' and 'Biscuits'. Adolescents ate 57·0 % and 51·3 % of non-core food at 'Eateries' or with 'Friends', compared with 33·2 % and 32·1 % at 'Home' or with 'Parents'. Persistent foods consumed across eating contexts were 'Regular soft drinks' and 'Chips & potato products'. Regular soft drinks contribute the most energy and are the most popular non-core food consumed by adolescents regardless of context, and represent a good target for interventions to reduce non-core food consumption.

  13. Drilling history core hole DC-6 Hanford, Washington

    International Nuclear Information System (INIS)

    1978-06-01

    Core hole DC-6 was completed in May 1978 by Boyles Brothers Drilling Company, Spokane, Washington, under subcontract to Fenix and Scisson, Inc. The hole was cored for the US Department of Energy and the Rockwell Hanford Operations' Basalt Waste Isolation Program. Fenix and Scisson, Inc. furnished the engineering, daily supervision of the core drilling activities, and geologic core logging for hole DC-6. Core hole DC-6 is located within the boundary of the Hanford Site at the old Hanford town site. The Hanford Site coordinates for DC-6 are North 54,127.17 feet and West 17,721.00 feet. The surface elevation is approximately 402 feet above sea level. The purpose of core hole DC-6 was to core drill vertically through the basalt and interbed units for stratigraphic depth determination and core collection and to provide a borehole for hydrologic testing. The total depth of core hole DC-6 was 4336 feet. Core recovery was 98.4% of the total footage cored

  14. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

    1994-08-01

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10 -7 /year

  15. Quality Assurance in the removal and transport of the TMI-2 core

    International Nuclear Information System (INIS)

    Hayes, G.R.; Marsden, J.F.

    1988-01-01

    EG ampersand G Idaho, acting on behalf of the US Department of Energy (DOE), is cooperating with the owner of the TMI-2 plant, General Public Utilities Nuclear (GPUN), in the removal and transport of the damaged TMI-2 core to the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Quality Assurance (QA) played an important role in the removal and transport of the damaged TMI-2 core. To illustrate, the authors have chosen to discuss some of the important quality assurance techniques utilized in the design, fabrication, acceptance, and use of the three different types of equipment; the core boring machine, the core debris canisters, and the transport casks. Rather than a thorough discussion of the QA aspects of each task, the authors have purposely chosen to present only the key applications of quality assurance principles and methodology unique to each piece of equipment. The intent of this approach is to effectively communicate the importance of ''task teamwork'' in QA

  16. Experimental evaluation on the damages of different drilling modes to tight sandstone reservoirs

    Directory of Open Access Journals (Sweden)

    Gao Li

    2017-07-01

    Full Text Available The damages of different drilling modes to reservoirs are different in types and degrees. In this paper, the geologic characteristics and types of such damages were analyzed. Then, based on the relationship between reservoir pressure and bottom hole flowing pressure corresponding to different drilling modes, the experimental procedures on reservoir damages in three drilling modes (e.g. gas drilling, liquid-based underbalanced drilling and overbalanced drilling were designed. Finally, damage simulation experiments were conducted on the tight sandstone reservoir cores of the Jurassic Ahe Fm in the Tarim Basin and Triassic Xujiahe Fm in the central Sichuan Basin. It is shown that the underbalanced drilling is beneficial to reservoir protection because of its less damage on reservoir permeability, but it is, to some extent, sensitive to the stress and the empirical formula of stress sensitivity coefficient is obtained; and that the overbalanced drilling has more reservoir damages due to the invasion of solid and liquid phases. After the water saturation of cores rises to the irreducible water saturation, the decline of gas logging permeability speeds up and the damage degree of water lock increases. It is concluded that the laboratory experiment results of reservoir damage are accordant with the reservoir damage characteristics in actual drilling conditions. Therefore, this method reflects accurately the reservoir damage characteristics and can be used as a new experimental evaluation method on reservoir damage in different drilling modes.

  17. A novel enzyme-based acidizing system: Matrix acidizing and drilling fluid damage removal

    Energy Technology Data Exchange (ETDEWEB)

    Harris, R.E.; McKay, D.M. [Cleansorb Limited, Surrey (United Kingdom); Moses, V. [King`s College, London (United Kingdom)

    1995-12-31

    A novel acidizing process is used to increase the permeability of carbonate rock cores in the laboratory and to remove drilling fluid damage from cores and wafers. Field results show the benefits of the technology as applied both to injector and producer wells.

  18. Dislocation core structures in Si-doped GaN

    International Nuclear Information System (INIS)

    Rhode, S. L.; Fu, W. Y.; Sahonta, S.-L.; Kappers, M. J.; Humphreys, C. J.; Horton, M. K.; Pennycook, T. J.; Dusane, R. O.; Moram, M. A.

    2015-01-01

    Aberration-corrected scanning transmission electron microscopy was used to investigate the core structures of threading dislocations in plan-view geometry of GaN films with a range of Si-doping levels and dislocation densities ranging between (5 ± 1) × 10 8  and (10 ± 1) × 10 9  cm −2 . All a-type (edge) dislocation core structures in all samples formed 5/7-atom ring core structures, whereas all (a + c)-type (mixed) dislocations formed either double 5/6-atom, dissociated 7/4/8/4/9-atom, or dissociated 7/4/8/4/8/4/9-atom core structures. This shows that Si-doping does not affect threading dislocation core structures in GaN. However, electron beam damage at 300 keV produces 4-atom ring structures for (a + c)-type cores in Si-doped GaN

  19. Dislocation core structures in Si-doped GaN

    Energy Technology Data Exchange (ETDEWEB)

    Rhode, S. L., E-mail: srhode@imperial.ac.uk; Fu, W. Y.; Sahonta, S.-L.; Kappers, M. J.; Humphreys, C. J. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Horton, M. K. [Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom); Pennycook, T. J. [SuperSTEM, STFC Daresbury Laboratories, Warrington WA4 4AD (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Dusane, R. O. [Department of Metallurgical Engineering and Materials Science, Indian Institute of Technology Bombay, Mumbai 400076 (India); Moram, M. A. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

    2015-12-14

    Aberration-corrected scanning transmission electron microscopy was used to investigate the core structures of threading dislocations in plan-view geometry of GaN films with a range of Si-doping levels and dislocation densities ranging between (5 ± 1) × 10{sup 8} and (10 ± 1) × 10{sup 9} cm{sup −2}. All a-type (edge) dislocation core structures in all samples formed 5/7-atom ring core structures, whereas all (a + c)-type (mixed) dislocations formed either double 5/6-atom, dissociated 7/4/8/4/9-atom, or dissociated 7/4/8/4/8/4/9-atom core structures. This shows that Si-doping does not affect threading dislocation core structures in GaN. However, electron beam damage at 300 keV produces 4-atom ring structures for (a + c)-type cores in Si-doped GaN.

  20. Fatigue-damage evolution and damage-induced reduction of critical current of a Nb3Al superconducting composite

    International Nuclear Information System (INIS)

    Ochiai, S; Sekino, F; Sawada, T; Ohno, H; Hojo, M; Tanaka, M; Okuda, H; Koganeya, M; Hayashi, K; Yamada, Y; Ayai, N; Watanabe, K

    2003-01-01

    We have studied the fatigue-damage mechanism of a Nb 3 Al superconducting composite at room temperature, and the influences of the fatigue damages introduced at room temperature on the critical current at 4.2 K and the residual strength at room temperature. The main (largest) fatigue crack arose first in the clad copper and then extended into the inner core with an increasing number of stress cycles. The cracking of the Nb 3 Al filaments in the core region occurred at a late stage (around 60-90% of the fatigue life). Once the fracture of the core occurred, it extended very quickly, resulting in a quick reduction in critical current and the residual strength with increasing stress cycles. Such a behaviour was accounted for by the crack growth calculated from the S-N curves (the relation of the maximum stress to the number of stress cycles at failure) combined with the Paris law. The size and distribution of the subcracks along the specimen length, and therefore the reduction in critical current of the region apart from the main crack, were dependent on the maximum stress level. The large subcracks causing fracture of the Nb 3 Al filaments were formed when the maximum stress was around 300-460 MPa, resulting in large reduction in critical current, but not when the maximum stress was outside such a stress range

  1. A fast alternative to core plug tests for optimising injection water salinity for EOR

    DEFF Research Database (Denmark)

    Hassenkam, Tue; Andersson, Martin Peter; Hilner, Emelie Kristin Margareta

    2014-01-01

    of the clays which would lead to permanent reservoir damage but evidence of effectiveness at moderate salinity would offer the opportunity to dispose of produced water. The goal is to define boundary conditions so injection water salinity is high enough to prevent reservoir damage and low enough to induce...... the low salinity effect while keeping costs and operational requirements at a minimum. Traditional core plug testing for optimising conditions has some limitations. Each test requires a fresh sample, core testing requires sophisticated and expensive equipment, and reliable core test data requires several...... experiments can be done relatively quickly on very little material, it gives the possibility of testing salinity response on samples from throughout a reservoir and for gathering statistics. Our approach provides a range of data that can be used to screen core plug testing conditions and to provide extra data...

  2. Refractory metal component technology for in-core sensor design

    International Nuclear Information System (INIS)

    Cannon, C.P.

    1986-02-01

    Within recent years, an increasing concern over reactor safety has prompted tests that characterize reactor core environments during transient conditions. Such tests include the Loss-of-Fluid-Tests (Idaho National Engineering Lab (INEL)), Severe Fuel Damage Tests (INEL), Core Debris Rubble Tests (Sandia National Laboratories (SNL)), and similar tests performed by foreign nations. The in-core sensors for these tests require refractory metal components to be compatible with electrical insulator materials as well as materials comprising highly corrosive service mediums. This paper presents the refractory metal technology utilized to provide basic sensor designs in the above mentioned reactor tests

  3. Damage Control Technology - A Literature Review

    Science.gov (United States)

    2006-03-01

    The Canadian Navy has identified the reduction of the total operating cost ( TOC ) of new ships as a priority. The major contributors to the TOC of a...Corporation, California, USA AC-CAS Group Co. Ltd., Bangkok, Thailand Apollo Fire Detectors, Hempshire, England, UK Compania Panamena de Sistemas ...National Defence DRDC Defence Research and Development Canada TOC Total Operating Cost BDCS Battle Damage Control System DC-ARM Damage Control

  4. Interpretation of the Haestholmen in situ state of stress based on core damage observations

    International Nuclear Information System (INIS)

    Hakala, M.

    2000-01-01

    At the Haestholmen investigation site, direct in situ stress measurements, overcoring and hydraulic fracturing have been unsuccessful because of ring disking and horizontal hydraulic fracturing. Prior to this study, a detailed study on both core disking and ring disking was made, and based on those results an in situ state of stress interpretation method was developed. In this work this method is applied to the Haestholmen site. The interpretation is based on disk fracture type, spacing and shape. Also, the Hoek-Brown strength envelope and Poisson's ratio of intact rock are needed. The interpretation result is most reliable if both core disking and ring disking information at the same depth levels is available. A detailed core logging showed that ring disking is systematic below the -365 m level in the vertical overcoring stress measurement hole, HH-KR6. On the other hand, no representative core disking exists except for two points in two differently oriented subvertical boreholes HH-KR2 and HHKR7. Because the interpretation has to be based on ring disking only, upper and lower estimates for the vertical stress were set. These were gravitational and 67% of gravitational. Furthermore, the in situ stress state was assumed to be in horizontal and vertical planes, because the disking in vertical borehole HH-KR6 was not inclined. The interpretation resulted in a good estimate for the major horizontal stress but none of the horizontal stress rations ( 0.25, 0.5, 0.75 and 1.0 ) or vertical stress assumptions studied are clearly more probable the others. At the 500 m level the resulting maximum horizontal stress is 41 MPa. If a linear fit through the zero depth and zero stress point is applied, the maximum horizontal stress gradient is 0.0818 z MPa/m with a standard deviation between 5 and 12 per cent. The orientation of the major horizontal stress is 108 with standard deviation of 21 degrees. The interpreted major horizontal stress state also indicated that systematic

  5. The game damages on agricultural crops in Croatia

    Directory of Open Access Journals (Sweden)

    Hrvoje Novosel

    2012-12-01

    Full Text Available Conflicts between wildlife and humans have been reported from all over the world, but in Croatia the extent and intensity of the conflict is increasing. Agricultural damage by game is a major concern for both agricultural and wildlife agencies at the national level. In this study 4,695 cases of game damage over a 4-year period were analysed. Results indicated that the total amount of economic impact on agriculture from game damage was significant. The distribution of payments shows that a majority of payments have lower economic value with an average amount of single payment being 477.08 €. The annual number of payments was found to have a negative correlation coefficient (-0.469 to the total payment amount for damages. According to the number of payments (68% of the number of payments and the payment amount (60% of total payment amount, the crop most often damaged was maize. Analysis of the data found that there was a negative growth trend of payment frequency and total payment amount for grape vineyards. The correlation between yearly number of payments and yearly production was not calculated for any crop. The high seasonal nature of payments was a determent of seasonal regression using a dummy variable regression (r2=0.93. A comparison of the monthly number of payments and monthly amounts is depicted by a time series using a seasonal line. The impact of wild boar damage on agriculture crops, in total, leads to the conclusion that this game species is a major problem. The results showed a specific subset of game damage in Croatia and, as such, it can be extrapolated to provide insight into the damage caused by wild boar in other countries.

  6. Metabolite damage and repair in metabolic engineering design

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jiayi; Jeffryes, James G.; Henry, Christopher S.; Bruner, Steven D.; Hanson, Andrew D.

    2017-11-01

    The necessarily sharp focus of metabolic engineering and metabolic synthetic biology on pathways and their fluxes has tended to divert attention from the damaging enzymatic and chemical side-reactions that pathway metabolites can undergo. Although historically overlooked and underappreciated, such metabolite damage reactions are now known to occur throughout metabolism and to generate (formerly enigmatic) peaks detected in metabolomics datasets. It is also now known that metabolite damage is often countered by dedicated repair enzymes that undo or prevent it. Metabolite damage and repair are highly relevant to engineered pathway design: metabolite damage reactions can reduce flux rates and product yields, and repair enzymes can provide robust, host-independent solutions. Herein, after introducing the core principles of metabolite damage and repair, we use case histories to document how damage and repair processes affect efficient operation of engineered pathways - particularly those that are heterologous, non-natural, or cell-free. We then review how metabolite damage reactions can be predicted, how repair reactions can be prospected, and how metabolite damage and repair can be built into genome-scale metabolic models. Lastly, we propose a versatile 'plug and play' set of well-characterized metabolite repair enzymes to solve metabolite damage problems known or likely to occur in metabolic engineering and synthetic biology projects.

  7. Destabilization of the Outer and Inner Mitochondrial Membranes by Core and Linker Histones

    Science.gov (United States)

    Cascone, Annunziata; Bruelle, Celine; Lindholm, Dan; Bernardi, Paolo; Eriksson, Ove

    2012-01-01

    Background Extensive DNA damage leads to apoptosis. Histones play a central role in DNA damage sensing and may mediate signals of genotoxic damage to cytosolic effectors including mitochondria. Methodology/Principal Findings We have investigated the effects of histones on mitochondrial function and membrane integrity. We demonstrate that both linker histone H1 and core histones H2A, H2B, H3, and H4 bind strongly to isolated mitochondria. All histones caused a rapid and massive release of the pro-apoptotic intermembrane space proteins cytochrome c and Smac/Diablo, indicating that they permeabilize the outer mitochondrial membrane. In addition, linker histone H1, but not core histones, permeabilized the inner membrane with a collapse of the membrane potential, release of pyridine nucleotides, and mitochondrial fragmentation. Conclusions We conclude that histones destabilize the mitochondrial membranes, a mechanism that may convey genotoxic signals to mitochondria and promote apoptosis following DNA damage. PMID:22523586

  8. In-core flow rate distribution measurement test of the JOYO irradiation core

    International Nuclear Information System (INIS)

    Suzuki, Toshihiro; Isozaki, Kazunori; Suzuki, Soju

    1996-01-01

    A flow rate distribution measurement test was carried out for the JOYO irradiation core (the MK-II core) after the 29th duty cycle operation. The main object of the test is to confirm the proper flow rate distribution at the final phase of the MK-II core. The each flow rate at the outlet of subassemblies was measured by the permanent magnetic flowmeter inserted avail of fuel exchange hole in the rotating plug. This is third test in the MK-II core, after 10 years absence from the final test (1985). Total of 550 subassemblies were exchanged and accumulated reactor operation time reached up to 38,000 hours from the previous test. As a conclusion, it confirmed that the flow rate distribution has been kept suitable in the final phase of the MK-II core. (author)

  9. AKT phosphorylates H3-threonine 45 to facilitate termination of gene transcription in response to DNA damage

    OpenAIRE

    Lee, Jong-Hyuk; Kang, Byung-Hee; Jang, Hyonchol; Kim, Tae Wan; Choi, Jinmi; Kwak, Sojung; Han, Jungwon; Cho, Eun-Jung; Youn, Hong-Duk

    2015-01-01

    Post-translational modifications of core histones affect various cellular processes, primarily through transcription. However, their relationship with the termination of transcription has remained largely unknown. In this study, we show that DNA damage-activated AKT phosphorylates threonine 45 of core histone H3 (H3-T45). By genome-wide chromatin immunoprecipitation sequencing (ChIP-seq) analysis, H3-T45 phosphorylation was distributed throughout DNA damage-responsive gene loci, particularly ...

  10. Damage correlation in theory and practice

    International Nuclear Information System (INIS)

    Doran, D.G.; Odette, G.R.; Simons, R.L.; Mansur, L.K.

    1977-01-01

    Common to all reactor development work is the problem of differences between the irradiation environments used for materials testing and those typical of service conditions. Efforts are being made to develop damage models that incorporate irradiation parameters such as type and energy of radiation, flux, and exposure. Models relating radiation damage production and microstructural evolution to changes in mechanical properties are primitive. Nevertheless, they suggest that the inability to account quantitatively for differences in test and service neutron spectra leads to overly conservative design of out-of-core components. Direct experimental corroboration is difficult because of the low neutron fluxes associated with the desired soft spectra. Further development of mechanistic models and new approaches to model testing are needed. Models of the growth stage of swelling, on the other hand, are relatively advanced. These models are discussed briefly as an example of how damage models can be used to help guide and analyze irradiation experiments. Accelerated damage studies using charged particles are expected to continue. Current empirical correlations of damage rates can be given a firmer theoretical basis as analysis of experiments and modeling of damage continue to improve. Damage correlation methodology practices in reactor design must necessarily follow different rules from that practiced in materials research and development. Nevertheless, decreasing the gap between them is a laudable objective with potentially significant economic impact

  11. Evaluation of re-criticality potential in Fukushima Dai-ichi reactors following core damage accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The re-criticality potential of the debris-bed, formed of the degraded core materials, cannot be ruled out during the cooling-down procedure of the Fukushima Dai-ichi NPPs. In this study the re-criticality potential has systematically investigated based on the core disruption phase analysis using a IMPACT-SAMPSON code prepared by The Institute of Applied Energy (IAE). The results obtained for the re-criticality potential, characterized by the eigen-values k-eff dependent on the debris composition formed at the core, RPV bottom, and PCV pedestal, are reflected to the arguments on the re-criticality prevention measures, such as timing and concentration of boron-compounds, during the cooling-down process of the Fukushima Dai-ichi NPPs. (author)

  12. Mechanisms of formation damage in matrix-permeability geothermal wells

    Energy Technology Data Exchange (ETDEWEB)

    Bergosh, J.L.; Wiggins, R.B.; Enniss, D.O.

    1982-04-01

    Tests were conducted to determine mechanisms of formation damage that can occur in matrix permeability geothermal wells. Two types of cores were used in the testing, actual cores from the East Mesa Well 78-30RD and cores from a fairly uniform generic sandstone formation. Three different types of tests were run. The East Mesa cores were used in the testing of the sensitivity of core to filtrate chemistry. The tests began with the cores exposed to simulated East Mesa brine and then different filtrates were introduced and the effects of the fluid contrast on core permeability were measured. The East Mesa cores were also used in the second series of tests which tested formation sandstone cores were used in the third test series which investigated the effects of different sizes of entrained particles in the fluid. Tests were run with both single-particle sizes and distributions of particle mixes. In addition to the testing, core preparation techniques for simulating fracture permeability were evaluated. Three different fracture formation mechanisms were identified and compared. Measurement techniques for measuring fracture size and permeability were also developed.

  13. Modularized Functions of the Fanconi Anemia Core Complex

    Directory of Open Access Journals (Sweden)

    Yaling Huang

    2014-06-01

    Full Text Available The Fanconi anemia (FA core complex provides the essential E3 ligase function for spatially defined FANCD2 ubiquitination and FA pathway activation. Of the seven FA gene products forming the core complex, FANCL possesses a RING domain with demonstrated E3 ligase activity. The other six components do not have clearly defined roles. Through epistasis analyses, we identify three functional modules in the FA core complex: a catalytic module consisting of FANCL, FANCB, and FAAP100 is absolutely required for the E3 ligase function, and the FANCA-FANCG-FAAP20 and the FANCC-FANCE-FANCF modules provide nonredundant and ancillary functions that help the catalytic module bind chromatin or sites of DNA damage. Disruption of the catalytic module causes complete loss of the core complex function, whereas loss of any ancillary module component does not. Our work reveals the roles of several FA gene products with previously undefined functions and a modularized assembly of the FA core complex.

  14. The potential of permeability damage during the thermal recovery of the Cold Lake bitumen

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Wiwchar, B.; Gunter, W. D. [Alberta Research Council, Devon, AB (Canada); Dudley, J. S. [Imperial Oil Ltd., Sarnia, ON (Canada). Research Dept.

    1997-08-01

    It has been suggested that hydrothermal reactions of clay minerals, present in all oil sands deposits in the Clearwater Formation at Cold Lake, may cause permeability damage during thermal recovery. To gain an idea of the extent of the damage, two corefloods were conducted at 250 degrees C. The first period of permeability damage occurred during and shortly after the core was heated to 250 degrees C, the second period was a gradual process , but resulted in 65 per cent and 78 per cent respectively, whereas the third period occurred when fresh water was injected into the core. These periods of damage were attributed to thermally activated grain crushing and fines migration, hydrothermal reactions, and osmotic swelling of the hydrothermal clay, respectively. Laboratory results do not agree with field experiments, although there is some field evidence for the disruption of berthierine (a form of clay) grain coats and permeability damage due to subsequent fines migration. In view of this evidence it was suggested that injection wells should not be placed in berthierine-rich zone. 15 refs., 2 tabs., 7 figs.

  15. Core status computing system

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki.

    1982-01-01

    Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

  16. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    International Nuclear Information System (INIS)

    Guzina, Bojan; Kunerth, Dennis

    2014-01-01

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  17. Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-30

    A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

  18. Utilization of control rod drive (CRD) system for long term core cooling

    International Nuclear Information System (INIS)

    Huerta B, A.

    1991-01-01

    In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

  19. Metabolite damage and repair in metabolic engineering design.

    Science.gov (United States)

    Sun, Jiayi; Jeffryes, James G; Henry, Christopher S; Bruner, Steven D; Hanson, Andrew D

    2017-11-01

    The necessarily sharp focus of metabolic engineering and metabolic synthetic biology on pathways and their fluxes has tended to divert attention from the damaging enzymatic and chemical side-reactions that pathway metabolites can undergo. Although historically overlooked and underappreciated, such metabolite damage reactions are now known to occur throughout metabolism and to generate (formerly enigmatic) peaks detected in metabolomics datasets. It is also now known that metabolite damage is often countered by dedicated repair enzymes that undo or prevent it. Metabolite damage and repair are highly relevant to engineered pathway design: metabolite damage reactions can reduce flux rates and product yields, and repair enzymes can provide robust, host-independent solutions. Herein, after introducing the core principles of metabolite damage and repair, we use case histories to document how damage and repair processes affect efficient operation of engineered pathways - particularly those that are heterologous, non-natural, or cell-free. We then review how metabolite damage reactions can be predicted, how repair reactions can be prospected, and how metabolite damage and repair can be built into genome-scale metabolic models. Lastly, we propose a versatile 'plug and play' set of well-characterized metabolite repair enzymes to solve metabolite damage problems known or likely to occur in metabolic engineering and synthetic biology projects. Copyright © 2017 International Metabolic Engineering Society. All rights reserved.

  20. Mitochondrial Respiration Is Reduced in Atherosclerosis, Promoting Necrotic Core Formation and Reducing Relative Fibrous Cap Thickness.

    Science.gov (United States)

    Yu, Emma P K; Reinhold, Johannes; Yu, Haixiang; Starks, Lakshi; Uryga, Anna K; Foote, Kirsty; Finigan, Alison; Figg, Nichola; Pung, Yuh-Fen; Logan, Angela; Murphy, Michael P; Bennett, Martin

    2017-12-01

    Mitochondrial DNA (mtDNA) damage is present in murine and human atherosclerotic plaques. However, whether endogenous levels of mtDNA damage are sufficient to cause mitochondrial dysfunction and whether decreasing mtDNA damage and improving mitochondrial respiration affects plaque burden or composition are unclear. We examined mitochondrial respiration in human atherosclerotic plaques and whether augmenting mitochondrial respiration affects atherogenesis. Human atherosclerotic plaques showed marked mitochondrial dysfunction, manifested as reduced mtDNA copy number and oxygen consumption rate in fibrous cap and core regions. Vascular smooth muscle cells derived from plaques showed impaired mitochondrial respiration, reduced complex I expression, and increased mitophagy, which was induced by oxidized low-density lipoprotein. Apolipoprotein E-deficient (ApoE -/- ) mice showed decreased mtDNA integrity and mitochondrial respiration, associated with increased mitochondrial reactive oxygen species. To determine whether alleviating mtDNA damage and increasing mitochondrial respiration affects atherogenesis, we studied ApoE -/- mice overexpressing the mitochondrial helicase Twinkle (Tw + /ApoE -/- ). Tw + /ApoE -/- mice showed increased mtDNA integrity, copy number, respiratory complex abundance, and respiration. Tw + /ApoE -/- mice had decreased necrotic core and increased fibrous cap areas, and Tw + /ApoE -/- bone marrow transplantation also reduced core areas. Twinkle increased vascular smooth muscle cell mtDNA integrity and respiration. Twinkle also promoted vascular smooth muscle cell proliferation and protected both vascular smooth muscle cells and macrophages from oxidative stress-induced apoptosis. Endogenous mtDNA damage in mouse and human atherosclerosis is associated with significantly reduced mitochondrial respiration. Reducing mtDNA damage and increasing mitochondrial respiration decrease necrotic core and increase fibrous cap areas independently of changes in

  1. The influence of selenium status on body composition, oxidative DNA damage and total antioxidant capacity in newly diagnosed type 2 diabetes mellitus: A case-control study.

    Science.gov (United States)

    Othman, Fatimah Binti; Mohamed, Hamid Jan Bin Jan; Sirajudeen, K N S; Noh, Mohd Fairulnizal B Md; Rajab, Nor Fadilah

    2017-09-01

    Selenium is involved in the complex system of defense against oxidative stress in diabetes through its biological function of selenoproteins and the antioxidant enzyme. A case-control study was carried out to determine the association of plasma selenium with oxidative stress and body composition status presented in Type 2 Diabetes Mellitus (T2DM) patient and healthy control. This study involved 82 newly diagnosed T2DM patients and 82 healthy controls. Plasma selenium status was determined with Graphite Furnace Atomic Absorption Spectrometry. Body Mass Index, total body fat and visceral fat was assessed for body composition using Body Composition Analyzer (TANITA). Oxidative DNA damage and total antioxidant capacity were determined for oxidative stress biomarker status. In age, gender and BMI adjustment, no significant difference of plasma selenium level between T2DM and healthy controls was observed. There was as a significant difference of Oxidative DNA damage and total antioxidant capacity between T2DM patients and healthy controls with tail DNA% 20.62 [95% CI: 19.71,21.49] (T2DM), 17.67 [95% CI: 16.87,18.56] (control); log tail moment 0.41[95% CI: 0.30,0.52] (T2DM), 0.41[95% CI: 0.30,0.52] (control); total antioxidant capacity 0.56 [95% CI: 0.54,0.58] (T2DM), 0.60 [95% CI: 0.57,0.62] (control). Waist circumference, BMI, visceral fat, body fat and oxidative DNA damage in the T2DM group were significantly lower in the first plasma selenium tertile (38.65-80.90μg/L) compared to the second (80.91-98.20μg/L) and the third selenium tertiles (98.21-158.20μg/L). A similar trend, but not statistically significant, was observed in the control group. Copyright © 2016 Elsevier GmbH. All rights reserved.

  2. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  3. Guide to diagnosis and appraisal of AAR damage to concrete in structures

    CERN Document Server

    Rooij, Mario; Wood, Jonathan

    2013-01-01

    This book describes procedures and methodologies used predominantly to obtain a diagnosis of damaged concrete possibly caused by Alkali-Aggregate Reaction (AAR). It has two primary objectives, namely firstly to identify the presence of AAR reaction, and whether or not the reaction is the primary or contributory cause of damage in the concrete; and secondly, to establish its intensity (severity) in various members of a structure. It includes aspects such as field inspection of the structure, sampling, petrographic examination of core samples, and supplementary tests and analyses on cores, such as mechanical tests and chemical analysis. Evaluation of test data for prognosis, consequences and appraisal will be more fully set out in AAR-6.2.

  4. Fatigue Characterization of Fire Resistant Syntactic Foam Core Material

    Science.gov (United States)

    Hossain, Mohammad Mynul

    Eco-Core is a fire resistant material for sandwich structural application; it was developed at NC A&T State University. The Eco-Core is made of very small amount of phenolic resin and large volume of flyash by a syntactic process. The process development, static mechanical and fracture, fire and toxicity safety and water absorption properties and the design of sandwich structural panels with Eco-Core material was established and published in the literature. One of the important properties that is needed for application in transportation vehicles is the fatigue performance under different stress states. Fatigue data are not available even for general syntactic foams. The objective of this research is to investigate the fatigue performance of Eco-Core under three types of stress states, namely, cyclic compression, shear and flexure, then document failure modes, and develop empherical equations for predicting fatigue life of Eco-Core under three stress states. Compression-Compression fatigue was performed directly on Eco-Core cylindrical specimen, whereas shear and flexure fatigue tests were performed using sandwich beam made of E glass-Vinyl Ester face sheet and Eco-Core material. Compression-compression fatigue test study was conducted at two values of stress ratios (R=10 and 5), for the maximum compression stress (sigmamin) range of 60% to 90% of compression strength (sigmac = 19.6 +/- 0.25 MPa) for R=10 and 95% to 80% of compression strength for R=5. The failure modes were characterized by the material compliance change: On-set (2% compliance change), propagation (5%) and ultimate failure (7%). The number of load cycles correspond to each of these three damages were characterized as on-set, propagation and total lives. A similar approach was used in shear and flexure fatigue tests with stress ratio of R=0.1. The fatigue stress-number of load cycles data followed the standard power law equation for all three stress states. The constant of the equation were

  5. X-ray atomic scattering factors of low-Z ions with a core hole

    International Nuclear Information System (INIS)

    Hau-Riege, Stefan P.

    2007-01-01

    Short and intense x-ray pulses may be used for atomic-resolution diffraction imaging of single biological molecules. One of the dominant damage mechanisms is atomic ionization, resulting in a large fraction of atoms with core holes. We calculated the atomic scattering factor of atoms with atomic charge numbers between 3 and 10 in different ionization states with and without a core hole. Our results show that orbital occupation and the change of the orbitals upon core ionization (core relaxation) have a significant impact on the diffraction pattern

  6. Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

    1986-01-01

    As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

  7. Discussion on the re-irradiated fuel assembly with damaged guide vanes

    International Nuclear Information System (INIS)

    Li Ligang

    2013-01-01

    In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

  8. Protective role of Hippopahe leaves against kidney damage in total body 60Co-gamma-irradiated mice

    International Nuclear Information System (INIS)

    Saini, Manu; Prasad, Jagdish; Bala, Madhu

    2012-01-01

    Hippophae rhamnoides has diverse therapeutic applications in Indian, Chinese and Tibetan medicine. Our earlier studies have shown that pretreatment with Hippophae leaf extract rendered >90% survival in mice population. The objective of this study was to investigate the role of our herbal preparation from Hippophae leaf against radiation induced kidney damage. The study was conducted with Strain 'A' male mice weighing approximately 28 ±2 g. The mice were administered Hippophae leaf extract, 30 minutes prior to 60 cobalt-gamma irradiation. The weight of kidneys and histological changes in kidney tissues at the light microscopic level were examined at 2, 5, 7, 10 and 15 days after treatment. The results showed that no significant change was observed in kidney weight after 60 cobalt-gamma irradiation. The glomerular damage in the form of glomerular sclerosis and percentage of damaged glomeruli; tubular damage in form of tubular dilations; apoptosis, and interstitial hemorrhages in renal cortex was also observed after radiation treatment. The pretreatment with Hippophae leaf extract countered most of the histological alterations induced by radiation. In comparison to radiation alone group, there was a significant decrease (p 60 cobalt gamma radiation induced damage. (author)

  9. Total body irradiation: current indications; L`irradiation corporelle totale: les indications actuelles

    Energy Technology Data Exchange (ETDEWEB)

    Giraud, P.; Danhier, S.; Dubray, B.; Cosset, J.M. [Institut Curie, 75 - Paris (France)

    1998-05-01

    The choice of dose and fractionation for total body irradiation is made difficult by the large number of considerations to be taken into account. The outcome of bone marrow transplantation after total body irradiation can be understood in terms of tumor cell killing, engraftment, and normal tissue damage, each of these endpoints being influenced by irradiation-, disease-, transplant-, and patient- related factors. Interpretation of clinical data is further hampered by the overwhelming influence of logistic constraints, the small numbers of randomized studies, and the concomitant variations in total dose and fraction size or dose rate. So far, three cautious conclusions can be drawn in order to tentatively adapt the total body irradiation schedule to clinically-relevant situations. Firstly, the organs at risk for normal tissue damage (lung, liver, lens, kidney) are protected by delivering small doses per fraction at low dose rate. This suggests that, when toxicity is at stake (e.g. in children), fractionated irradiation should be preferred, provided that inter-fraction intervals are long enough. Secondly, fractionated irradiation should be avoided in case of T-cell depleted transplant, given the high risk of graft rejection in this setting. An alternative would be to increase total (or fractional) dose of fractionated total body irradiation, but this approach is likely to induce more normal tissue toxicity. Thirdly, clinical data have shown higher relapse rates in chronic myeloid leukemia after fractionated or low dose rate total body irradiation, suggesting that fractionated irradiation should not be recommended, unless total (or fractional) dose is increased. Total body irradiation-containing regimens, primarily cyclophosphamide / total body irradiation, are either equivalent to or better than the chemotherapy-only regimens, primarily busulfan / cyclophosphamide. Busulfan / cyclophosphamide certainly represents a reasonable alternative, especially in patients who

  10. Disaster and Contingency Planning for Scientific Shared Resource Cores.

    Science.gov (United States)

    Mische, Sheenah; Wilkerson, Amy

    2016-04-01

    Progress in biomedical research is largely driven by improvements, innovations, and breakthroughs in technology, accelerating the research process, and an increasingly complex collaboration of both clinical and basic science. This increasing sophistication has driven the need for centralized shared resource cores ("cores") to serve the scientific community. From a biomedical research enterprise perspective, centralized resource cores are essential to increased scientific, operational, and cost effectiveness; however, the concentration of instrumentation and resources in the cores may render them highly vulnerable to damage from severe weather and other disasters. As such, protection of these assets and the ability to recover from a disaster is increasingly critical to the mission and success of the institution. Therefore, cores should develop and implement both disaster and business continuity plans and be an integral part of the institution's overall plans. Here we provide an overview of key elements required for core disaster and business continuity plans, guidance, and tools for developing these plans, and real-life lessons learned at a large research institution in the aftermath of Superstorm Sandy.

  11. Effect of varying geometrical parameters of trapezoidal corrugated-core sandwich structure

    Directory of Open Access Journals (Sweden)

    Zaid N.Z.M.

    2017-01-01

    Full Text Available Sandwich structure is an attractive alternative that increasingly used in the transportation and aerospace industry. Corrugated-core with trapezoidal shape allows enhancing the damage resistance to the sandwich structure, but on the other hand, it changes the structural response of the sandwich structure. The aim of this paper is to study the effect of varying geometrical parameters of trapezoidal corrugated-core sandwich structure under compression loading. The corrugated-core specimen was fabricated using press technique, following the shape of trapezoidal shape. Two different materials were used in the study, glass fibre reinforced plastic (GFRP and carbon fibre reinforced plastic (CFRP. The result shows that the mechanical properties of the core in compression loading are sensitive to the variation of a number of unit cells and the core thickness.

  12. Nuclear damage compensation and energy reform

    International Nuclear Information System (INIS)

    Yokemoto, Masafumi

    2013-01-01

    Nuclear damage compensation and energy reform were closely related. Nuclear damage compensation cost should be part of generation cost of nuclear power. Extend of nuclear damage compensation was limited by compensation standard of Tokyo Electric Power Co. (TEPCO) following guidelines of Dispute Reconciliation Committee for Nuclear Damage Compensation. TEPCO had already paid compensation of about two trillion yen until now, which was only a part of total damage compensation cost. TEPCO had been provided more than 3.4 trillion yen by Nuclear Damage Liability Facilitation Cooperation, which would be put back by nuclear operators including TEPCO. TEPCO could obtain present raising funds and try to reconstruct business with restart of nuclear power, which might disturb energy reform. Present nuclear damage compensation scheme had better be reformed with learning more from Minamata disease case in Japan. (T. Tanaka)

  13. [Damage control in trauma patients with hemodynamic instability].

    Science.gov (United States)

    Müller, Thorben; Doll, Dietrich; Kliebe, Frank; Ruchholtz, Steffen; Kühne, Christian

    2010-10-01

    The term "Damage-control" is borrowed from naval terminology. It means the initial control of a damaged ship. Because of the lethal triad in multiple injured patients the classical concept of definitive surgically therapy in the acute phase of the injury has a high rate of complications such as exsanguination, sepsis, heart failure and multiple organ failure. The core idea of the damage control concept was to minimize the additional trauma by surgical operations in these critical patients in the first phase. This means temporary control of a hemorrhage and measures for stopping abdominal contamination. After 24 - 48 hours in the intensive care unit and correction of physiological disturbances further interventions are performed for definitively treatment of the injuries. Summarized, the damage control strategy comprises an abbreviated operation, intensive care unit resuscitation, and a return to the operating room for the definitive operation after hemodynamic stabilisation of the patient. © Georg Thieme Verlag Stuttgart · New York.

  14. New insights regarding ATWS for BWRS

    International Nuclear Information System (INIS)

    Drouin, M.T.; Kolaczkowski, A.M.; LaChance, J.L.; Ferrell, W.L.

    1987-01-01

    Anticipated transients without scram (ATWS) accident sequences have been found in past studies to have a relatively high core damage frequency (ranging from 5.4E-6 to 3E-4 per year) that represents a significant contribution to the total core damage frequency (ranging from 7-to-33%). Results of analyses for the two boiling water reactors (BWRs) analyzed as part of NUREG/CR-4550 indicate both a lower core damage frequency (ranging from 2E-7 to 1E-6 per year) and a lower contribution to the total core damage frequency (ranging from <1-to-10%). Based on these updated analyses, newer insights on the effects of reactor power equilibration, recirculation pump trip, high and low pressure injection and high pressure seal failure coupled with a detailed accident sequence analysis have resulted in lowering the significance of ATWS to core damage frequency

  15. Plastic damage induced fracture behaviors of dental ceramic layer structures subjected to monotonic load.

    Science.gov (United States)

    Wang, Raorao; Lu, Chenglin; Arola, Dwayne; Zhang, Dongsheng

    2013-08-01

    The aim of this study was to compare failure modes and fracture strength of ceramic structures using a combination of experimental and numerical methods. Twelve specimens with flat layer structures were fabricated from two types of ceramic systems (IPS e.max ceram/e.max press-CP and Vita VM9/Lava zirconia-VZ) and subjected to monotonic load to fracture with a tungsten carbide sphere. Digital image correlation (DIC) and fractography technology were used to analyze fracture behaviors of specimens. Numerical simulation was also applied to analyze the stress distribution in these two types of dental ceramics. Quasi-plastic damage occurred beneath the indenter in porcelain in all cases. In general, the fracture strength of VZ specimens was greater than that of CP specimens. The crack initiation loads of VZ and CP were determined as 958 ± 50 N and 724 ± 36 N, respectively. Cracks were induced by plastic damage and were subsequently driven by tensile stress at the elastic/plastic boundary and extended downward toward to the veneer/core interface from the observation of DIC at the specimen surface. Cracks penetrated into e.max press core, which led to a serious bulk fracture in CP crowns, while in VZ specimens, cracks were deflected and extended along the porcelain/zirconia core interface without penetration into the zirconia core. The rupture loads for VZ and CP ceramics were determined as 1150 ± 170 N and 857 ± 66 N, respectively. Quasi-plastic deformation (damage) is responsible for crack initiation within porcelain in both types of crowns. Due to the intrinsic mechanical properties, the fracture behaviors of these two types of ceramics are different. The zirconia core with high strength and high elastic modulus has better resistance to fracture than the e.max core. © 2013 by the American College of Prosthodontists.

  16. TMI-2 Core Shipping Preparations

    International Nuclear Information System (INIS)

    Ball, L.J.; Barkanic, R.J.; Conaway, W.T. II; Schmoker, D. S.; Post, Roy G.

    1988-01-01

    Shipping the damaged core from the Unit 2 reactor of Three Mile Island Nuclear Power Station near Harrisburg, PA, to the Idaho National Engineering Laboratory near Idaho Falls, ID, required development and implementation of a completely new spent fuel transportation system. This paper describes of the equipment developed, the planning and activities used to implement the hard-ware-systems into the facilities, and the planning involved in making the rail shipments. It also includes a summary of recommendations resulting from this experience. (author)

  17. Deep ice coring at Dome Fuji Station, Antarctica

    Directory of Open Access Journals (Sweden)

    Yoshiyuki Fujii

    1999-03-01

    Full Text Available Deep ice coring was carried out at Dome Fuji Station, Antarctica in 1995 and 1996 following a pilot borehole drilled and cased with FRP pipes in 1993,and reached 2503.52m in December 1996. Total numbers of ice coring runs below the pilot borehole and chip collection were 1369 and 837 respectively. The mean coring depths per run and per day were 1.75m and 8.21m respectively. We report the outline of the coring operation, the system, coring method, and troubles encountered during the coring work.

  18. The total triterpenoid saponins of Xanthoceras sorbifolia improve learning and memory impairments through against oxidative stress and synaptic damage.

    Science.gov (United States)

    Ji, Xue-Fei; Chi, Tian-Yan; Liu, Peng; Li, Lu-Yi; Xu, Ji-Kai; Xu, Qian; Zou, Li-Bo; Meng, Da-Li

    2017-02-15

    X. sorbifolia is a widely cultivated ecologicalcrop in the north of China which is used to produce biodiesel fuel. It also possesses special medicinal value and has attracted keen interests of researchers to explore its bioactivity. To extract the total triterpenoid saponins from the husk of X. sorbifolia (TSX) and investigate its effects on Alzheimer's disease (AD). TSX was prepared via modern extraction techniques. Its effects on two AD animal models, as well as the preliminary mechanism were investigated comprehensively. The behavioral experiments including Y maze test, Morris water maze test and passive avoidance test were performed to observe the learning and memory abilities of the animals. ELISA assays, transmission electron microscope observation and Western blotting were employed in mechanism study. TSX, the main composition of X. sorbifolia, accounted for 88.77% in the plant material. It could significantly increase the spontaneous alternation in Y maze test (F (6, 65)=3.209, Plearning and memory. The preliminary mechanism might associate with its protection effects against oxidative stress damage, cholinergic system deficiency and synaptic damage. TSX are perfectly suitable for AD patients as medicine or functional food, which would be a new candidate to treat AD. Copyright © 2016 Elsevier GmbH. All rights reserved.

  19. Fracture Characterization of Sandwich Face/Core Interfaces

    DEFF Research Database (Denmark)

    Manca, Marcello

    of load transfer between the faces and the core layer is lost, the debonds are considered as primary damage initiators. Under fatigue loading the debonds may evolve into cracks that cause a reduction in structural performance and consequent failure. At present most structural design is based on “life-time...... of sandwich structures is defects that are introduced in the manufacturing process. It is inevitable that areas of the face sheets will not fully adhere to the core resulting in defects known as “debonds”. Debonds can also be induced in-service due to e.g. localised impact loading or overloading. As the means...... such result it is important to devise new experimental and analytical techniques to establish the multi-mode fracture characteristics of sandwich plate structures and accordingly develop methods to inhibit defect propagation. This thesis deals with characterization of fracture between face and core...

  20. Toxicity of sediment cores collected from the Ashtabula River in northeastern Ohio, USA, to the amphipod Hyalella azteca

    Science.gov (United States)

    Ingersoll, C.G.; Kemble, N.E.; Kunz, J.L.; Brumbaugh, W.G.; MacDonald, D.D.; Smorong, D.

    2009-01-01

    This study was conducted to support a Natural Resource Damage Assessment and Restoration project associated with the Ashtabula River in Ohio. The objective of the study was to evaluate the chemistry and toxicity of 50 sediment samples obtained from five cores collected from the Ashtabula River (10 samples/core, with each 10-cm-diameter core collected to a total depth of about 150 cm). Effects of chemicals of potential concern (COPCs) measured in the sediment samples were evaluated by measuring whole-sediment chemistry and whole-sediment toxicity in the sediment samples (including polycyclic aromatic hydrocarbons [PAHs], polychlorinated biphenyls [PCBs], organochlorine pesticides, and metals). Effects on the amphipod Hyalella azteca at the end of a 28-day sediment toxicity test were determined by comparing survival or length of amphipods in individual sediment samples in the cores to the range of responses of amphipods exposed to selected reference sediments that were also collected from the cores. Mean survival or length of amphipods was below the lower limit of the reference envelope in 56% of the sediment samples. Concentrations of total PCBs alone in some samples or concentrations of total PAHs alone in other samples were likely high enough to have caused the reduced survival or length of amphipods (i.e., concentrations of PAHs or PCBs exceeded mechanistically based and empirically based sediment quality guidelines). While elevated concentrations of ammonia in pore water may have contributed to the reduced length of amphipods, it is unlikely that the reduced length was caused solely by elevated ammonia (i.e., concentrations of ammonia were not significantly correlated with the concentrations of PCBs or PAHs and concentrations of ammonia were elevated both in the reference sediments and in the test sediments). Results of this study show that PAHs, PCBs, and ammonia are the primary COPCs that are likely causing or substantially contributing to the toxicity to

  1. Thawing of lithium in the SP-100 reactor core configuration

    International Nuclear Information System (INIS)

    Magee, P.M.; Malovrh, J.W.; REineking, W.H.

    1986-01-01

    The General Electric SP-100 Liquid Metal Reactor is designed to be launched with the lithium coolant in the reactor and primary loops frozen. Initial startup of the system in space, after a satisfactory orbit is achieved, will be accomplished by slowly increasing the power in the reactor core and using the heat generated to melt the lithium, first in the reactor, and then progressively down the primary loops. This technique significantly facilitates ground handling, reduces vibrational loads during vehicle launch and minimized the shuttle bay heat load. The challenge is to thaw the coolant and startup the system within an acceptable time without structural damage. The test results clearly demonstrate that thawing of the lithium in the SP-100 reactor core can be done rapidly without structural damage and, thus, support the selected concept of SP-100 launch with frozen lithium and thaw/startup in space

  2. Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Johnson, Richard; Schultz, Richard

    2010-01-01

    The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

  3. Damage cost of the Dan River coal ash spill

    International Nuclear Information System (INIS)

    Dennis Lemly, A.

    2015-01-01

    The recent coal ash spill on the Dan River in North Carolina, USA has caused several negative effects on the environment and the public. In this analysis, I report a monetized value for these effects after the first 6 months following the spill. The combined cost of ecological damage, recreational impacts, effects on human health and consumptive use, and esthetic value losses totals $295,485,000. Because the environmental impact and associated economic costs of riverine coal ash spills can be long-term, on the order of years or even decades, this 6-month assessment should be viewed as a short-term preview. The total cumulative damage cost from the Dan River coal ash spill could go much higher. - Highlights: • Six-month post-spill damage cost exceeded $295,000,000. • Components of cost include ecological, recreational, human health, property, and aesthetic values. • Attempts by the electric utility to “clean” the river left over 95% of coal ash behind. • Long-term impacts will likely drive the total damage cost much higher. - Damage costs of the Dan River coal ash spill are extensive and growing. The 6-month cost of that spill is valued at $295,485,000, and the long-term total cost is likely to rise substantially

  4. Damage visualization and deformation measurement in glass laminates during projectile penetration

    Directory of Open Access Journals (Sweden)

    Elmar Strassburger

    2014-06-01

    Full Text Available Transparent armor consists of glass-polymer laminates in most cases. The formation and propagation of damage in the different glass layers has a strong influence on the ballistic resistance of such laminates. In order to clarify the course of events during projectile penetration, an experimental technique was developed, which allows visualizing the onset and propagation of damage in each single layer of the laminate. A telecentric objective lens was used together with a microsecond video camera that allows recording 100 frames at a maximum rate of 1 MHz in a backlit photography set-up. With this technique, the damage evolution could be visualized in glass laminates consisting of four glass layers with lateral dimensions 500 mm × 500 mm. Damage evolution was recorded during penetration of 7.62 mm AP projectiles with tungsten carbide core and a total mass of 11.1 g in the impact velocity range from 800 to 880 m/s. In order to measure the deformation of single glass plates within the laminates, a piece of reflecting tape was attached to the corresponding glass plate, and photonic Doppler velocimetry (PDV was applied. With the photonic Doppler velocimeter, an infrared laser is used to illuminate an object to be measured and the Doppler-shifted light is superimposed to a reference light beam at the detector. The simultaneous visualization and PDV measurement of the glass deformation allow determining the deformation at the time of the onset of fracture. The analysis of the experimental data was supported by numerical simulations, using the AUTODYN commercial hydro-code.

  5. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal fires during mid-loop operations. Volume 3, Part 1, Main report

    International Nuclear Information System (INIS)

    Musicki, Z.; Chu, T.L.; Yang, J.; Ho, V.; Hou, Y.M.; Lin, J.; Siu, N.

    1994-07-01

    During l989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than fun power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in ' the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few. procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

  6. Correlation of the concentration of the carbon-associated radiation damage levels with the total carbon concentration in silicon

    Energy Technology Data Exchange (ETDEWEB)

    Ferenczi, G.; Londos, C.A.; Pavelka, T.; Somogyi, M.; Mertens, A.

    1988-01-01

    The dominant carbon-related radiation damage center in silicon was studied in detail by deep level transient spectroscopy. Samples with different carbon and oxygen content were implanted with gradually increasing proton fluence. Two energetically closely spaced levels were revealed and tentative identities were assigned. One at E/sub T/+E/sub V/ = 0.344 eV (sigma/sub p/ = 1.1 x 10/sup -16/ cm/sup 2/) is assigned as the C+O/sub i/ complex, and that at E/sub T/+E/sub V/ = 0.370 eV (sigma/sub p/ = 8 x 10/sup -18/ cm/sup 2/) is assigned as the C/sub s/-Si/sub i/-C/sub s/ complex. It was shown that the concentration of these defects is correlated to the total concentration of carbon in the crystal.

  7. Nonlinear Ultrasonic Diagnosis and Prognosis of ASR Damage in Dry Cask Storage

    International Nuclear Information System (INIS)

    Qu, Jianmin; Bazant, Zdenek; Jacobs, Laurence; Guimaraes, Maria

    2015-01-01

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

  8. Nonlinear Ultrasonic Diagnosis and Prognosis of ASR Damage in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Jianmin [Northwestern Univ., Evanston, IL (United States); Bazant, Zdenek [Northwestern Univ., Evanston, IL (United States); Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Guimaraes, Maria [Electrical Power Research Institute, Palo Alto, CA (United States)

    2015-11-30

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

  9. Studies on the strategies of minimizing radiation damage

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Hee Yong; Sohn, Young Sook

    1998-04-01

    We studied on the strategies of minimizing radiation damage in animal system. To this end we studied following areas of research (1) mechanisms involved in bone marrow damage after total body irradiation, (2) extraction of components that are useful in protecting hematopoietic system from radiation damage, (3) cell therapy approach in restoring the damaged tissue, (4) development of radioprotective chemical reagent, and (5) epidemiological study on the population that had been exposed to radiation.

  10. Studies on the strategies of minimizing radiation damage

    International Nuclear Information System (INIS)

    Chung, Hee Yong; Sohn, Young Sook

    1998-04-01

    We studied on the strategies of minimizing radiation damage in animal system. To this end we studied following areas of research 1) mechanisms involved in bone marrow damage after total body irradiation, 2) extraction of components that are useful in protecting hematopoietic system from radiation damage, 3) cell therapy approach in restoring the damaged tissue, 4) development of radioprotective chemical reagent, and 5) epidemiological study on the population that had been exposed to radiation

  11. Proposition of a core model for the thorium molten salt reactor (TMSR) minimizing the graphite moderator quantity in core; Proposition d'un modele de coeur pour le RSF thorium minimisant la quantite de moderateur graphite en coeur

    Energy Technology Data Exchange (ETDEWEB)

    Nuttin, A

    2004-07-01

    This work deals with the problem of fast damage of graphite in the core of TMSR. The approach consists to minimize the quantity of graphite used in the core (by an increase of the voluminal power) and then to extract and to reprocess. (O.M.)

  12. Phosphate, carbonate and organic matter distribution in sediment cores off Bombay-Saurashtra coast, India

    Digital Repository Service at National Institute of Oceanography (India)

    Setty, M.G.A.P.; Rao, Ch.M.

    relationship. The total phosphorus value is high in the core from the slope, almost uniform in the cores from the outer shelf and a core from the nearby basin, but poor in a core (no. 3) considered to be from a 'closed basin' within the shelf. The total...

  13. Preparations to receive and store the TMI-2 core debris

    International Nuclear Information System (INIS)

    Ayers, A.L.R. Jr.; Lilburn, B.J. Jr.

    1986-01-01

    The March 1979 accident at Unit 2 of Three Mile Island Nuclear Power Station (TMI-2) resulted in considerable damage to the core of the reactor. The core debris will be packaged in canisters and transported by rail cask to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. A significant part of recovering from the TMI-2 accident involves receiving and storing the TMI-2 core debris canisters at INEL. This paper highlights preparations for receiving the rail cask at INEL, unloading canisters from the cask in the Hot Shop of Test Area North Building 607, and storing/monitoring those canisters in the Water Pit for up to 30 years

  14. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  15. Compact multipurpose sub-sampling and processing of in-situ cores with press (pressurized core sub-sampling and extrusion system)

    Energy Technology Data Exchange (ETDEWEB)

    Anders, E.; Muller, W.H. [Technical Univ. of Berlin, Berlin (Germany). Chair of Continuum Mechanics and Material Theory

    2008-07-01

    Climate change, declining resources and over-consumption result in a need for sustainable resource allocation, habitat conservation and claim for new technologies and prospects for damage-containment. In order to increase knowledge of the environment and to define potential hazards, it is necessary to get an understanding of the deep biosphere. In addition, the benthic conditions of sediment structure and gas hydrates, temperature, pressure and bio-geochemistry must be maintained during the sequences of sampling, retrieval, transfer, storage and downstream analysis. In order to investigate highly instable gas hydrates, which decomposes under pressure and temperature change, a suite of research technologies have been developed by the Technische Universitat Berlin (TUB), Germany. This includes the pressurized core sub-sampling and extrusion system (PRESS) that was developed in the European Union project called HYACE/HYACINTH. The project enabled well-defined sectioning and transfer of drilled pressure-cores obtained by a rotary corer and fugro pressure corer into transportation and investigation chambers. This paper described HYACINTH pressure coring and the HYACINTH core transfer. Autoclave coring tools and HYACINTH core logging, coring tools, and sub-sampling were also discussed. It was concluded that possible future applications include, but were not limited to, research in shales and other tight formations, carbon dioxide sequestration, oil and gas exploration, coalbed methane, and microbiology of the deep biosphere. To meet the corresponding requirements and to incorporate the experiences from previous expeditions, the pressure coring system would need to be redesigned to adapt it to the new applications. 3 refs., 5 figs.

  16. Constrained core solutions for totally positive games with ordered players

    NARCIS (Netherlands)

    van den Brink, J.R.; van der Laan, G.; Vasil'ev, V.

    2014-01-01

    In many applications of cooperative game theory to economic allocation problems, such as river-, polluted river- and sequencing games, the game is totally positive (i.e., all dividends are nonnegative), and there is some ordering on the set of the players. A totally positive game has a nonempty

  17. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  18. Disaster and Contingency Planning for Scientific Shared Resource Cores

    Science.gov (United States)

    Wilkerson, Amy

    2016-01-01

    Progress in biomedical research is largely driven by improvements, innovations, and breakthroughs in technology, accelerating the research process, and an increasingly complex collaboration of both clinical and basic science. This increasing sophistication has driven the need for centralized shared resource cores (“cores”) to serve the scientific community. From a biomedical research enterprise perspective, centralized resource cores are essential to increased scientific, operational, and cost effectiveness; however, the concentration of instrumentation and resources in the cores may render them highly vulnerable to damage from severe weather and other disasters. As such, protection of these assets and the ability to recover from a disaster is increasingly critical to the mission and success of the institution. Therefore, cores should develop and implement both disaster and business continuity plans and be an integral part of the institution’s overall plans. Here we provide an overview of key elements required for core disaster and business continuity plans, guidance, and tools for developing these plans, and real-life lessons learned at a large research institution in the aftermath of Superstorm Sandy. PMID:26848285

  19. Historical summary of the Three Mile Island Unit 2 core debris transportation campaign

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, R.C.; Tyacke, M.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Quinn, G.J. [Wastren, Inc., Germantown, MD (United States)

    1993-03-01

    Transport of the damaged core materials from the Unit 2 reactor of the Three Mile Island Nuclear Power Station (TMI-2) to the Idaho National Engineering Laboratory (INEL) for examination and storage presented many technical and institutional challenges, including assessing the ability to transport the damaged core; removing and packaging core debris in ways suitable for transport; developing a transport package that could both meet Federal regulations and interface with the facilities at TMI-2 and the INEL; and developing a transport plan, support logistics, and public communications channels suited to the task. This report is a historical summary of how the US Department of Energy addressed those challenges and transported, received, and stored the TMI-2 core debris at the INEL. Subjects discussed include preparations for transport, loading at TMI-2, institutional issues, transport operations, receipt and storage at the INEL, governmental inquiries/investigations, and lessons learned. Because of public attention focused on the TMI-2 Core Debris Transport Program, the exchange of information between the program and public was extensive. This exchange is a focus for parts of this report to explain why various operations were conducted as they were and why certain technical approaches were employed. And, because of that exchange, the program may have contributed to a better public understanding of such actions and may contribute to planning and execution of similar future actions.

  20. Historical summary of the Three Mile Island Unit 2 core debris transportation campaign

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Tyacke, M.J.; Quinn, G.J.

    1993-03-01

    Transport of the damaged core materials from the Unit 2 reactor of the Three Mile Island Nuclear Power Station (TMI-2) to the Idaho National Engineering Laboratory (INEL) for examination and storage presented many technical and institutional challenges, including assessing the ability to transport the damaged core; removing and packaging core debris in ways suitable for transport; developing a transport package that could both meet Federal regulations and interface with the facilities at TMI-2 and the INEL; and developing a transport plan, support logistics, and public communications channels suited to the task. This report is a historical summary of how the US Department of Energy addressed those challenges and transported, received, and stored the TMI-2 core debris at the INEL. Subjects discussed include preparations for transport, loading at TMI-2, institutional issues, transport operations, receipt and storage at the INEL, governmental inquiries/investigations, and lessons learned. Because of public attention focused on the TMI-2 Core Debris Transport Program, the exchange of information between the program and public was extensive. This exchange is a focus for parts of this report to explain why various operations were conducted as they were and why certain technical approaches were employed. And, because of that exchange, the program may have contributed to a better public understanding of such actions and may contribute to planning and execution of similar future actions

  1. Supplemental materials for the ICDP-USGS Eyreville A, B, and C core holes, Chesapeake Bay impact structure: Core-box photographs, coring-run tables, and depth-conversion files

    Science.gov (United States)

    Durand, C.T.; Edwards, L.E.; Malinconico, M.L.; Powars, D.S.

    2009-01-01

    During 2005-2006, the International Continental Scientific Drilling Program and the U.S. Geological Survey drilled three continuous core holes into the Chesapeake Bay impact structure to a total depth of 1766.3 m. A collection of supplemental materials that presents a record of the core recovery and measurement data for the Eyreville cores is available on CD-ROM at the end of this volume and in the GSA Data Repository. The supplemental materials on the CD-ROM include digital photographs of each core box from the three core holes, tables of the three coring-run logs, as recorded on site, and a set of depth-conversion programs. In this chapter, the contents, purposes, and basic applications of the supplemental materials are briefly described. With this information, users can quickly decide if the materials will apply to their specific research needs. ?? 2009 The Geological Society of America.

  2. Tunable engineered skin mechanics via coaxial electrospun fiber core diameter.

    Science.gov (United States)

    Blackstone, Britani Nicole; Drexler, Jason William; Powell, Heather Megan

    2014-10-01

    Autologous engineered skin (ES) offers promise as a treatment for massive full thickness burns. Unfortunately, ES is orders of magnitude weaker than normal human skin causing it to be difficult to apply surgically and subject to damage by mechanical shear in the early phases of engraftment. In addition, no manufacturing strategy has been developed to tune ES biomechanics to approximate the native biomechanics at different anatomic locations. To enhance and tune ES biomechanics, a coaxial (CoA) electrospun scaffold platform was developed from polycaprolactone (PCL, core) and gelatin (shell). The ability of the coaxial fiber core diameter to control both scaffold and tissue mechanics was investigated along with the ability of the gelatin shell to facilitate cell adhesion and skin development compared to pure gelatin, pure PCL, and a gelatin-PCL blended fiber scaffold. CoA ES exhibited increased cellular adhesion and metabolism versus PCL alone or gelatin-PCL blend and promoted the development of well stratified skin with a dense dermal layer and a differentiated epidermal layer. Biomechanics of the scaffold and ES scaled linearly with core diameter suggesting that this scaffold platform could be utilized to tailor ES mechanics for their intended grafting site and reduce graft damage in vitro and in vivo.

  3. Tunable Engineered Skin Mechanics via Coaxial Electrospun Fiber Core Diameter

    Science.gov (United States)

    Blackstone, Britani Nicole; Drexler, Jason William

    2014-01-01

    Autologous engineered skin (ES) offers promise as a treatment for massive full thickness burns. Unfortunately, ES is orders of magnitude weaker than normal human skin causing it to be difficult to apply surgically and subject to damage by mechanical shear in the early phases of engraftment. In addition, no manufacturing strategy has been developed to tune ES biomechanics to approximate the native biomechanics at different anatomic locations. To enhance and tune ES biomechanics, a coaxial (CoA) electrospun scaffold platform was developed from polycaprolactone (PCL, core) and gelatin (shell). The ability of the coaxial fiber core diameter to control both scaffold and tissue mechanics was investigated along with the ability of the gelatin shell to facilitate cell adhesion and skin development compared to pure gelatin, pure PCL, and a gelatin-PCL blended fiber scaffold. CoA ES exhibited increased cellular adhesion and metabolism versus PCL alone or gelatin-PCL blend and promoted the development of well stratified skin with a dense dermal layer and a differentiated epidermal layer. Biomechanics of the scaffold and ES scaled linearly with core diameter suggesting that this scaffold platform could be utilized to tailor ES mechanics for their intended grafting site and reduce graft damage in vitro and in vivo. PMID:24712409

  4. Radiation damage studies on new liquid scintillators and liquid-core scintillating fibers

    International Nuclear Information System (INIS)

    Golovkin, S.V.

    1994-01-01

    The radiation resistant of some new liquid scintillation and capillaries filled with liquid scintillators has been presented. It was found that scintillation efficiency of the scintillator based on 1-methyl naphthalene with a new R39 only by 10% at the dose of 190 Mrad and the radiation resistance of thin liquid-core scintillating was decreased fibers exceeded 60 Mrad. 35 refs

  5. No post-no core approach to restore severely damaged posterior teeth: An up to 10-year retrospective study of documented endocrown cases.

    Science.gov (United States)

    Belleflamme, Marcia M; Geerts, Sabine O; Louwette, Marie M; Grenade, Charlotte F; Vanheusden, Alain J; Mainjot, Amélie K

    2017-08-01

    The objectives of the present study were to (1) retrospectively evaluate documented cases of ceramic and composite endocrowns performed using immediate dentin sealing (IDS); (2) correlate failures with clinical parameters such as tooth preparation characteristics and occlusal parameters. 99 documented cases of endocrowns were evaluated after a mean observation period of 44.7±34.6months. A classification of restorations was established in function of the level of damage of residual tooth tissues after preparation, from 1 to 3. Evaluation was performed according to FDI criteria and endodontic outcomes were analyzed. Occlusal risk factors were examined and fractographic analysis was performed in case of fracture. 48.4% of patients were shown to present occlusal risk factors. 75.8% of restorations were Class 3 endocrowns. 56.6% were performed on molars, 41.4% on premolars and 2.0% on canines. 84.8% were performed in lithium-disilicate glass-ceramic and 12.1% in Polymer-Infiltrated Ceramic Network (PICN) material. The survival and success rates of endocrowns were 99.0% and 89.9% respectively, while the 10-year Kaplan-Meier estimated survival and success rates were 98.8% and 54.9% respectively. Ten failures were detected: periodontal disease (n=3), endocrown debonding (n=2), minor chipping (n=2), caries recurrence (n=2) and major fractures (n=1). Due to the reduced amount of failures, no statistical correlation could be established with clinical parameters. Endocrowns were shown to constitute a reliable approach to restore severely damaged molars and premolars, even in the presence of extensive coronal tissue loss or occlusal risk factors, such as bruxism or unfavorable occlusal relationships. Practitioners should consider the endocrown instead of the post and core approach to restore severely damaged non-vital posterior teeth. This minimally invasive solution reduces the risk of catastrophic failures and is easily performed. The use of IDS procedure and lithium

  6. Contribution to the microchemistry of smoke damage by fluoride. The migration of fluorides in plant tissue. 2. The visible damage

    Energy Technology Data Exchange (ETDEWEB)

    Reckendorfer, P

    1953-01-01

    In continuation of former investigations, a theory of damage caused by fluorine compounds on green plants was developed. It is possible to differentiate between acute and chronic damages by use of microanalytical estimation of total fluorine and inorganic and organic fluorine compounds in the plants.

  7. ESFR core optimization and uncertainty studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

    2015-01-01

    In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

  8. Method of determination of thermo-acoustic coolant instability boundaries in reactor core at NPPs with WWER

    International Nuclear Information System (INIS)

    Skalozubov, Volodymyr; Kolykhanov, Viktor; Kovryzhkin, Yuriy

    2007-01-01

    The regulatory body of Ukraine, the National Atomic Energy Company and the Scientific and Production Centre have led team-works concerned with previously unstudied factors or phenomena affecting reactor safety. As a result it is determined that the thermo-acoustic coolant instability conditions can appear in the core at definite operating WWER regimes. Considerable cyclic dynamic loads affect fuel claddings over thermo-acoustic pressure oscillations. These loads can result in inadmissible cassette design damage and containment damage. Taking into account calculation and experimental research authors submit a method of on-line assessment of WWER core state concerning thermo-acoustic coolant instability. According to this method, the thermo-acoustic coolant instability appearance conditions can be estimated using normal registered parameters (pressure, temperature, heat demand etc.). At operative modes, a WWER-1000 core is stable to tracheotomies oscillations, but reduction of coolant discharge through the core for some times can result in thermo-acoustic coolant instability. Thermo-acoustic instability appears at separate transitional modes concerned with reactor scram and unloading/loading at all power units. When thermo-acoustic instability begins in transitional modes, core elements are under influence of high-frequency coolant pressure pulsations for a long time (tens of hours)

  9. Coal transportation road damage

    International Nuclear Information System (INIS)

    Burtraw, D.; Harrison, K.; Pawlowski, J.A.

    1994-01-01

    Heavy trucks are primarily responsible for pavement damage to the nation's highways. In this paper we evaluate the pavement damage caused by coal trucks. We analyze the chief source of pavement damage (vehicle weight per axle, not total vehicle weight) and the chief cost involved (the periodic overlay that is required when a road's surface becomes worn). This analysis is presented in two stages. In the first section we present a synopsis of current economic theory including simple versions of the formulas that can be: used to calculate costs of pavement wear. In the second section we apply this theory to a specific example proximate to the reference environment for the Fuel Cycle Study in New Mexico in order to provide a numerical measure of the magnitude of the costs

  10. Calculation of the Incremental Conditional Core Damage Probability on the Extension of Allowed Outage Time

    International Nuclear Information System (INIS)

    Kang, Dae Il; Han, Sang Hoon

    2006-01-01

    RG 1.177 requires that the conditional risk (incremental conditional core damage probability and incremental conditional large early release probability: ICCDP and ICLERP), given that a specific component is out of service (OOS), be quantified for a permanent change of the allowed outage time (AOT) of a safety system. An AOT is the length of time that a particular component or system is permitted to be OOS while the plant is operating. The ICCDP is defined as: ICCDP = [(conditional CDF with the subject equipment OOS)- (baseline CDF with nominal expected equipment unavailabilities)] [duration of the single AOT under consideration]. Any event enabling the component OOS can initiate the time clock for the limiting condition of operation for a nuclear power plant. Thus, the largest ICCDP among the ICCDPs estimated from any occurrence of the basic events for the component fault tree should be selected for determining whether the AOT can be extended or not. If the component is under a preventive maintenance, the conditional risk can be straightforwardly calculated without changing the CCF probability. The main concern is the estimations of the CCF probability because there are the possibilities of the failures of other similar components due to the same root causes. The quantifications of the risk, given that a subject equipment is in a failed state, are performed by setting the identified event of subject equipment to TRUE. The CCF probabilities are also changed according to the identified failure cause. In the previous studies, however, the ICCDP was quantified with the consideration of the possibility of a simultaneous occurrence of two CCF events. Based on the above, we derived the formulas of the CCF probabilities for the cases where a specific component is in a failed state and we presented sample calculation results of the ICCDP for the low pressure safety injection system (LPSIS) of Ulchin Unit 3

  11. Economic damage caused by a nuclear reactor accident

    International Nuclear Information System (INIS)

    Goemans, T.; Schwarz, J.J.

    1988-01-01

    This study is directed towards the estimation of the economic damage which arises from a severe possible accident with a newly built 1000 MWE nuclear power plant in the Netherlands. A number of cases have been considered which are specified by the weather conditions during and the severity of the accident and the location of the nuclear power plant. For each accident case the economic damage has been estimated for the following impact categories: loss of the power plant, public health, evacuation and relocation of population, export of agricultural products, working and living in contaminated regions, decontamination, costs of transportation and incoming foreign tourism. The consequences for drinking water could not be quantified adequately. The total economic damage could reach 30 billion guilders. Besides the power plant itself, loss of export and decreasing incoming foreign tourism determine an important part of the total damage. 12 figs.; 52 tabs

  12. Parameters affecting of Akkuyu's safety assessment for severe core damages

    Science.gov (United States)

    Kavun, Yusuf; Karasulu, Muzaffer

    2015-07-01

    We have looked at all past core meltdowns (Three Mile Island, Chernobyl and Fukushima incidents) and postulated the fourth one might be taking place in the future most probably in a newly built reactors anywhere of the earth in any type of NPP. The probability of this observation is high considering the nature of the machine and human interaction. Operation experience is a very significant parameter as well as the safety culture of the host nation. The concerns is not just a lack of experience with industry with the new comers, but also the infrastructure and established institutions who will be dealing with the Emergencies. Lack of trained and educated Emergency Response Organizations (ERO) is a major concern. The culture on simple fire drills even makes the difference when a severe condition occurs in the industry. The study assumes the fourth event will be taking place at the Akkuyu NGS and works backwards as required by the "what went wrong " scenarios and comes up with interesting results. The differences studied in depth to determine the impact to the severe accidents. The all four design have now core catchers. We have looked at the operator errors'like in TMI); Operator errors combined with design deficiencies(like in Chernobyl) and natural disasters( like in Fukushima) and found operator errors to be more probable event on the Akkuyu's postulated next incident. With respect to experiences of the operators we do not have any data except for long and successful operating history of the Soviet design reactors up until the Chernobyl incident. Since the Akkuyu will be built, own and operated by the Russians we have found no alarming concerns at the moment. At the moment, there is no body be able to operate those units in Turkey. Turkey is planning to build the required manpower during the transition period. The resolution of the observed parameters lies to work and educate, train of the host nation and exercise together.

  13. Critical Evolution of Damage Toward System-Size Failure in Crystalline Rock

    Science.gov (United States)

    Renard, François; Weiss, Jérôme; Mathiesen, Joachim; Ben-Zion, Yehuda; Kandula, Neelima; Cordonnier, Benoît

    2018-02-01

    Rock failure under shear loading conditions controls earthquake and faulting phenomena. We study the dynamics of microscale damage precursory to shear faulting in a quartz-monzonite rock representative of crystalline rocks of the continental crust. Using a triaxial rig that is transparent to X-rays, we image the mechanical evolution of centimeter-size core samples by in situ synchrotron microtomography with a resolution of 6.5 μm. Time-lapse three-dimensional images of the samples inside the rig provide a unique data set of microstructural evolution toward faulting. Above a yield point there is a gradual weakening during which microfractures nucleate and grow until this damage span the whole sample. This leads to shear faults oriented about 30° to the main compressive stress in agreement with Anderson's theory and macroscopic failure. The microfractures can be extracted from the three-dimensional images, and their dynamics and morphology (i.e., number, volume, orientation, shape, and largest cluster) are quantified as a function of increasing stress toward failure. The experimental data show for the first time that the total volume of microfractures, the rate of damage growth, and the size of the largest microfracture all increase and diverge when approaching faulting. The average flatness of the microfractures (i.e., the ratio between the second and third eigenvalues of their covariance matrix) shows a significant decrease near failure. The precursors to faulting developing in the future faulting zone are controlled by the evolving microfracture population. Their divergent dynamics toward failure is reminiscent of a dynamical critical transition.

  14. Transliterating transmission of genome damage in rats

    International Nuclear Information System (INIS)

    Slovinska, L.; Sanova, S.; Misurova, E.

    2004-01-01

    We studied the influence of gamma radiation (3 Gy) on slowly proliferating liver tissue of male rats and their progeny considering to induction and duration of latent damage. The irradiation caused latent cytogenetic damage in the liver in irradiated males of the F 0 generation manifesting itself during induced proliferation of hepatocytes (after partial hepatectomy) by reduced proliferating activity, a higher frequency of chromosomal aberrations and higher proportion of cells with apoptotic DNA fragments. In the progeny of irradiated males (F 1 and F 2 generation), the latent genome damage manifested itself during liver regeneration after partial hepatectomy by similar, but less pronounced changes compared with irradiated males of the parental generation. This finding indicates the transfer of the part of radiation-induced genome damage from parents to their progeny. Irradiation of F 1 and F 2 progeny of irradiated males (their total radiation load was 3+3 Gy, 3+0+3 Gy respectively) caused less changes as irradiation of progeny of non-irradiated control males (their total radiation load was 0+3 Gy, 0+0+3 Gy respectively). (authors)

  15. Phototherapy causes DNA damage in peripheral mononuclear leukocytes in term infants.

    Science.gov (United States)

    Aycicek, Ali; Kocyigit, Abdurrahim; Erel, Ozcan; Senturk, Hakan

    2008-01-01

    Our aim was to determine whether endogenous mononuclear leukocyte DNA strand is a target of phototherapy. The study included 65 term infants aged between 3-10 days that had been exposed to intensive (n = 23) or conventional (n = 23) phototherapy for at least 48 hours due to neonatal jaundice, and a control group (n = 19). DNA damage was assayed by single-cell alkaline gel electrophoresis (comet assay). Plasma total antioxidant capacity and total oxidant status levels were also measured, and correlation between DNA damage and oxidative stress was investigated. Mean values of DNA damage scores in both the intensive and conventional phototherapy groups were significantly higher than those in the control group (p Total oxidant status levels in both the intensive and conventional phototherapy groups were significantly higher than those in the control group (p = 0.005). Mean (standard deviation) values were 18.1 (4.2), 16.9 (4.4), 13.5 (4.2) micromol H2O2 equivalent/L, respectively. Similarly, oxidative stress index levels in both the intensive and conventional phototherapy groups were significantly higher than those in the control group (p = 0.041). Plasma total antioxidant capacity and total bilirubin levels did not differ between the groups (p > 0.05). There were no significant correlations between DNA damage scores and bilirubin, total oxidant status and oxidative stress levels in either phototherapy group (p > 0.05). Both conventional phototherapy and intensive phototherapy cause endogenous mononuclear leukocyte DNA damage in jaundiced term infants.

  16. Influence of alkali-silica reaction and crack orientation on the uniaxial compressive strength of concrete cores from slab bridges

    DEFF Research Database (Denmark)

    Antonio Barbosa, Ricardo; Gustenhoff Hansen, Søren; Hansen, Kurt Kielsgaard

    2018-01-01

    ASR-damaged flat slab bridges in service. Furthermore, the influence of the ASR-induced crack orientation on the compressive strength and the Young’s modulus is investigated. Uniaxial compression tests, visual observations, and thin section examinations were performed on more than 100 cores drilled...... from the three severely ASR-damaged flat slab bridges. It was found that the orientation of ASR-induced cracks has a significant influence on the uniaxial compressive strength and the stress-strain relationship of the tested cores. The compressive strength in a direction parallel to ASR cracks can...

  17. Effect of low dose rate irradiation on doped silica core optical fibers

    International Nuclear Information System (INIS)

    Friebele, E.J.; Askins, C.G.; Gingerich, M.E.

    1984-01-01

    The optical attenuation induced in multimode doped silica core optical fiber waveguides by a year's exposure to low dose rate (1 rad/day) ionizing radiation was studied, allowing a characterization of fibers deployed in these environments and a determination of the permanent induced loss in the waveguides. Variations in the induced attenuation at 0.85 μm have been observed with changes in the dose rate between 1 rad/day and 9000 rads/min. These dose rate dependences have been found to derive directly from the recovery that occurs during the exposure; the recovery data predict little or no dose rate dependence of the damage at 1.3 μm. The low dose rate exposure has been found to induce significant permanent attenuation in the 0.7-1.7-μm spectral region in all fibers containing P in the core, whether doped uniformly across the diameter or constrained to a narrow spike on the centerline. Whereas permanent loss was induced at 0.85 μm in a P-free binary Ge-doped silica core fiber by the year's exposure, virtually no damage was observed at 1.3 μm

  18. Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

    OpenAIRE

    Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

    2010-01-01

    Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

  19. Analysis and prevention of water hammer for the emergency core cooling system

    International Nuclear Information System (INIS)

    Zhao Jun

    2008-01-01

    Emergency core cooling system (ECCS) is an engineered safety feature of nuclear power plant. If the water hammer happens during ECCS injection, the piping system may be broken. It will cause loss of ECC system and affect the safety of reactor core. Based on the functions and characteristics of ECCS and the theory of water hammer, the paper analyzed the potential risk of water hammer in ECCS in Qinshan III, and proposed modifications to prevent the water-hammer damage during ECCS injection. (authors)

  20. Finite Element Modelling and Analysis of Damage Detection Methodology in Piezo Electric Sensor and Actuator Integrated Sandwich Cantilever Beam

    Science.gov (United States)

    Pradeep, K. R.; Thomas, A. M.; Basker, V. T.

    2018-03-01

    Structural health monitoring (SHM) is an essential component of futuristic civil, mechanical and aerospace structures. It detects the damages in system or give warning about the degradation of structure by evaluating performance parameters. This is achieved by the integration of sensors and actuators into the structure. Study of damage detection process in piezoelectric sensor and actuator integrated sandwich cantilever beam is carried out in this paper. Possible skin-core debond at the root of the cantilever beam is simulated and compared with undamaged case. The beam is actuated using piezoelectric actuators and performance differences are evaluated using Polyvinylidene fluoride (PVDF) sensors. The methodology utilized is the voltage/strain response of the damaged versus undamaged beam against transient actuation. Finite element model of piezo-beam is simulated in ANSYSTM using 8 noded coupled field element, with nodal degrees of freedoms are translations in the x, y directions and voltage. An aluminium sandwich beam with a length of 800mm, thickness of core 22.86mm and thickness of skin 0.3mm is considered. Skin-core debond is simulated in the model as unmerged nodes. Reduction in the fundamental frequency of the damaged beam is found to be negligible. But the voltage response of the PVDF sensor under transient excitation shows significantly visible change indicating the debond. Piezo electric based damage detection system is an effective tool for the damage detection of aerospace and civil structural system having inaccessible/critical locations and enables online monitoring possibilities as the power requirement is minimal.

  1. Damage detection in composite materials using Lamb wave methods

    Science.gov (United States)

    Kessler, Seth S.; Spearing, S. Mark; Soutis, Constantinos

    2002-04-01

    Cost-effective and reliable damage detection is critical for the utilization of composite materials. This paper presents part of an experimental and analytical survey of candidate methods for in situ damage detection of composite materials. Experimental results are presented for the application of Lamb wave techniques to quasi-isotropic graphite/epoxy test specimens containing representative damage modes, including delamination, transverse ply cracks and through-holes. Linear wave scans were performed on narrow laminated specimens and sandwich beams with various cores by monitoring the transmitted waves with piezoceramic sensors. Optimal actuator and sensor configurations were devised through experimentation, and various types of driving signal were explored. These experiments provided a procedure capable of easily and accurately determining the time of flight of a Lamb wave pulse between an actuator and sensor. Lamb wave techniques provide more information about damage presence and severity than previously tested methods (frequency response techniques), and provide the possibility of determining damage location due to their local response nature. These methods may prove suitable for structural health monitoring applications since they travel long distances and can be applied with conformable piezoelectric actuators and sensors that require little power.

  2. Probabilistic Fatigue Damage Program (FATIG)

    Science.gov (United States)

    Michalopoulos, Constantine

    2012-01-01

    FATIG computes fatigue damage/fatigue life using the stress rms (root mean square) value, the total number of cycles, and S-N curve parameters. The damage is computed by the following methods: (a) traditional method using Miner s rule with stress cycles determined from a Rayleigh distribution up to 3*sigma; and (b) classical fatigue damage formula involving the Gamma function, which is derived from the integral version of Miner's rule. The integration is carried out over all stress amplitudes. This software solves the problem of probabilistic fatigue damage using the integral form of the Palmgren-Miner rule. The software computes fatigue life using an approach involving all stress amplitudes, up to N*sigma, as specified by the user. It can be used in the design of structural components subjected to random dynamic loading, or by any stress analyst with minimal training for fatigue life estimates of structural components.

  3. Timing of the Three Mile Island Unit 2 core degradation as determined by forensic engineering

    International Nuclear Information System (INIS)

    Henrie, J.O.

    1988-01-01

    Unlike computer simulation of an event, forensic engineering is the evaluation of recorded data and damaged as well as surviving components after an event to determine progressive causes of the event. Such an evaluation of the 1979 Three Mile Island Unit 2 accident indicates that gas began accumulating in steam, generator A at 6:10, or 130 min into the accident and, therefore, fuel cladding ruptures and/or zirconium-water reactions began at that time. Zirconium oxidation/hydrogen generation rates were highest (∼70 kg of hydrogen per minute) during the core quench and collapse at 175 min. By 180 min, over 85% of the hydrogen generated by the zirconium-water reaction had been produced, and ∼400 kg of hydrogen had accumulated in the reactor coolant system. At that time, hydrogen concentrations at the steam/water interfaces in both steam generators approached 90%. By 203 min, the damaged reactor core had been reflooded and has not been uncovered since that time. Therefore, the core was completely under water at 225 min, when molten core material flowed into the lower head of the reactor vessel. 10 refs., 7 figs., 1 tab

  4. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  5. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  6. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  7. Study on core flow distribution of the reference core design Mark-III of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Satoh, Sadao; Arai, Taketoshi; Miyamoto, Yoshiaki; Hirano, Mitsumasa

    1977-01-01

    Concerning the coolant flow distribution between fuel channels and other flow paths in the core, designated as Reference Core Mark-III of the Multi-purpose Experimental Very High Temperature Reactor, thermal analysis has been made of the control rods and other steel structures around the core to find the coolant flow rates (bypass flow) necessary to cool them to their safe operating temperatures. Calculations showed that adequate cooling could be achieved in the Mark-III Core by the bypass flow of 8% of the total reactor coolant flow, 4% each for the control-rod channels and for other structures. The thermal and coolant flow design bases, including the assumption of a 10% bypass flow, were thus confirmed to first approximation. (auth.)

  8. Experimental study of the mechanical behaviour of pin reinforced foam core sandwich materials under shear load

    International Nuclear Information System (INIS)

    Dimassi, M A; Brauner, C; Herrmann, A S

    2016-01-01

    Sandwich structures with a lightweight closed cell hard foam core have the potential to be used in primary structures of commercial aircrafts. Compared to honeycomb core sandwich, the closed cell foam core sandwich overcomes the issue of moisture take up and makes the manufacturing of low priced and highly integrated structures possible. However, lightweight foam core sandwich materials are prone to failure by localised external loads like low velocity impacts. Invisible cracks could grow in the foam core and threaten the integrity of the structure. In order to enhance the out-of-plane properties of foam core sandwich structures and to improve the damage tolerance (DT) dry fibre bundles are inserted in the foam core. The pins are infused with resin and co-cured with the dry fabric face sheets in an out-of-autoclave process. This study presents the results obtained from shear tests following DIN 53294-standard, on flat sandwich panels. All panels were manufactured with pin-reinforcement manufactured with the Tied Foam Core Technology (TFC) developed by Airbus. The effects of pin material (CFRP and GFRP) and pin volume fraction on the shear properties of the sandwich structure and the crack propagation were investigated and compared to a not pinned reference. It has been concluded that the pin volume fraction has a remarkable effect on the shear properties and damage tolerance of the observed structure. Increasing the pin volume fraction makes the effect of crack redirection more obvious and conserves the integrity of the structure after crack occurrence. (paper)

  9. Silica hollow core microstructured fibers for beam delivery in industrial and medical applications

    Directory of Open Access Journals (Sweden)

    Jonathan Dale Shephard

    2015-04-01

    Full Text Available The focus of this review is our recent work to develop microstructured hollow core fibers for two applications where the flexible delivery of a single mode beam is desired. Also, a review of other fiber based solutions is included.High power, short-pulsed lasers are widely used for micro-machining, providing high precision and high quality. However, the lack of truly flexible beam delivery systems limits their application to the processing of relatively small planar components. To address this, we developed hollow-core optical fibers for the 1 μm and green wavelength ranges. The hollow core overcomes the power delivery limitations of conventional silica fibers arising from nonlinear effects and material damage in the solid core. We have characterized such fibers in terms of power handling capability, damage threshold, bend loss and dispersion, and practically demonstrated delivery of high peak power pulses from the nanosecond to the femtosecond regime. Such fibers are ideal candidates for industrial laser machining applications.In laser surgical applications, meanwhile, an Er:YAG laser (2.94 μm is frequently the laser of choice because the water contained in tissue strongly absorbs this wavelength. If this laser beam is precisely delivered damage to surrounding tissue can be minimized. A common delivery method of surgical lasers, for use in the operating theatre, is articulated arms that are bulky, cumbersome and unsuitable for endoscopic procedures. To address this need for flexible mid-IR delivery we developed silica based hollow core fibers. By minimizing the overlap of the light with glass it is possible to overcome the material absorption limits of silica and achieve low attenuation. Additionally, it is possible to deliver pulse energies suitable for the ablation of both hard and soft tissue even with very small bend radii. The flexibility and small physical size of systems based on these fibers will enable new minimally invasive surgical

  10. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report

  11. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

  12. Report on damaged FLIP TRIGA fuel

    International Nuclear Information System (INIS)

    Feltz, Donald E.; Randall, John D.; Schumacher, Robert F.

    1977-01-01

    Damaged FLIP elements were discovered, positioned adjacent to the transient rod. It then became apparent that this was not the failure of a defective, element but a heretofore unknown operating or design problem. The damaged elements are described as having bulges in the cladding and unevenly spaced dark rings along the fuelled portion of the element. Possible causes are investigated, including: defective fuel elements, incorrectly calculated power distributions in the core and in the elements, water leakage into the void follower of the transient rod, and improper safety limit for FLIP fuel. Based on measurements and calculations that have been experimentally verified it is concluded that the safety limit was not exceeded or even closely approached. It is also concluded that the problem is due entirely due to some phenomena occurring during pulsing, and that the steady state history of the fuel is not a factor

  13. 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

  14. Current status of core needle biopsy of the thyroid

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jung Hwan [Dept. of Radiology and Research Institute of Radiology, Asan Medical Center, University of Ulsan College of Medicine, Seoul (Korea, Republic of)

    2017-04-15

    Thyroid nodules are a common clinical problem. Fine-needle aspiration (FNA) and large-needle biopsy have been used to diagnose thyroid nodules. Before the 1980s, large-needle biopsy was the standard procedure for the thyroid, but FNA became the standard diagnostic tool in the 1980s because it is a safe procedure that leads to accurate diagnoses. With advances in core needle biopsy (CNB) devices (i.e., spring-activated core needles) and development of high-resolution ultrasound, it has become possible to make accurate diagnoses while minimizing complications. Although 18- to 21-gauge core needles can be used to biopsy thyroid nodules, 18-gauge needles are most commonly used in Korea. The relationships among the size of the needle, the number of core specimens, and diagnostic accuracy have not yet been conclusively established, but the general tendency is that thinner needles cause less damage to the normal thyroid, but allow a smaller amount of thyroid tissue to be biopsied to be obtained. These relationships may be validated in the future.

  15. Storm damage in the Black Forest caused by the winter storm "Lothar" – Part 1: Airborne damage assessment

    Directory of Open Access Journals (Sweden)

    J. Schmoeckel

    2008-08-01

    Full Text Available An airborne survey of the Black Forest as affected by the winter storm "Lothar" in 1999 is performed by means of a color line scanner (CLS with a CCD sensor, whose data in a visible and a near-infrared channel provide the Normalized Difference Vegetation Index (NDVI as a measure of the damage in previously intact forest areas. The camera data, height data from a digital evelation model (DEM, land use information, and soil data are georeferenced and processed in a geographic information system (GIS to derive relationship of the damage pattern to the characteristics of the local orography and soil types. The data cover an area of 4900 km2, 2767 km2 of which were forested. The 363 detected storm damage areas with a minimum detection size of 1.5 ha amount to 0.8% of the total forest area. Visual inspections at certain sites prove that none of the larger damage areas are missed, but areas smaller than 1.5 ha cause the total damage area to be up to twice our result, i.e. ≈1.6% of the forest area. More than 50% of the detected damaged areas are smaller than 5 ha and most of them have a size ranging from 1.5 to 3.5 ha. Forests on slopes with an inclination angle between 10 and 15 degrees show the highest fraction of damaged forest, doubling those on plains and below 5 degrees inclination angle. Forests on northwestern slopes are more affected than those on southwestern and western slopes, which faced the wind during highest wind speed occurrence. In contrast to other studies, this paper shows, that in steep areas, lee slopes are more damaged than the luv slopes. As expected, wet to moist soils represent an unstable location for the trees. But also medium-dry to dry locations that were considered to be relatively stable exhibited a highly damaged forest fraction. This can be attributed to mostly saturated soil from previous rain.

  16. First combined total reflection X-ray fluorescence and grazing incidence X-ray absorption spectroscopy characterization of aeolian dust archived in Antarctica and Alpine deep ice cores

    Energy Technology Data Exchange (ETDEWEB)

    Cibin, G. [Diamond Light Source, Harwell Science and Innovation Campus, Didcot, Oxon OX110DE (United Kingdom); IMONT/EIM, Ente Italiano della Montagna, P.za dei Caprettari 70, 00176 Roma (Italy); Universita' degli Studi di Roma Tre, Dipartimento di Scienze Geologiche, L.go S. Leonardo Murialdo 1, 00146 Roma (Italy)], E-mail: giannantonio.cibin@diamond.ac.uk; Marcelli, A. [INFN - Laboratori Nazionali di Frascati, P.O. Box 13, 00044 Frascati (Roma) (Italy); Maggi, V. [Universita degli Studi di Milano-Bicocca, Dipartimento di Scienze dell' Ambiente e del Territorio, Piazza della Scienza 1, 20126 Milano (Italy); Sala, M. [Universita degli Studi di Milano-Bicocca, Dipartimento di Scienze dell' Ambiente e del Territorio, Piazza della Scienza 1, 20126 Milano (Italy); Universita degli Studi di Milano, Dipartimento di Scienze della Terra ' A. Desio' , Sez. Mineralogia, Via Mangiagalli 34, 20133 Milano (Italy); Marino, F.; Delmonte, B. [Universita degli Studi di Milano-Bicocca, Dipartimento di Scienze dell' Ambiente e del Territorio, Piazza della Scienza 1, 20126 Milano (Italy); Albani, S. [Universita degli Studi di Milano-Bicocca, Dipartimento di Scienze dell' Ambiente e del Territorio, Piazza della Scienza 1, 20126 Milano (Italy); Universita degli Studi di Siena, Dottorato in Scienze Polari, via Laterina 8, 53100 Siena (Italy); Pignotti, S. [IMONT/EIM, Ente Italiano della Montagna, P.za dei Caprettari 70, 00176 Roma (Italy)

    2008-12-15

    Aeolian mineral dust archived in polar and mid latitude ice cores represents a precious proxy for assessing environmental and climatic variations at different timescales. In this respect, the identification of dust mineralogy plays a key role. In this work we performed the first preliminary X-ray absorption spectroscopy (XAS) experiments on mineral dust particles extracted from Antarctic and from Alpine firn cores using grazing incidence geometry at the Fe K-edge. A dedicated high vacuum experimental chamber was set up for normal-incidence and total-reflection X-Ray Fluorescence and Absorption Spectroscopy analyses on minor amounts of mineral materials at the Stanford Synchrotron Radiation Laboratory. Results show that this experimental technique and protocol allows recognizing iron inclusion mineral fraction on insoluble dust in the 1-10 {mu}g range.

  17. First combined total reflection X-ray fluorescence and grazing incidence X-ray absorption spectroscopy characterization of aeolian dust archived in Antarctica and Alpine deep ice cores

    International Nuclear Information System (INIS)

    Cibin, G.; Marcelli, A.; Maggi, V.; Sala, M.; Marino, F.; Delmonte, B.; Albani, S.; Pignotti, S.

    2008-01-01

    Aeolian mineral dust archived in polar and mid latitude ice cores represents a precious proxy for assessing environmental and climatic variations at different timescales. In this respect, the identification of dust mineralogy plays a key role. In this work we performed the first preliminary X-ray absorption spectroscopy (XAS) experiments on mineral dust particles extracted from Antarctic and from Alpine firn cores using grazing incidence geometry at the Fe K-edge. A dedicated high vacuum experimental chamber was set up for normal-incidence and total-reflection X-Ray Fluorescence and Absorption Spectroscopy analyses on minor amounts of mineral materials at the Stanford Synchrotron Radiation Laboratory. Results show that this experimental technique and protocol allows recognizing iron inclusion mineral fraction on insoluble dust in the 1-10 μg range

  18. Thermal Margin Calculation of the CAREM-25 Core

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    During the operation in steady state and anticipated operational transient of a nuclear reactor it is necessary to avoid the damage in the fuel elements induced by thermal or hydraulic effects.To satisfy that design bases safety limits are established and calculation methodologies are defined to verify them.In the particular case of the reactor CAREM-25 reactor where the core is cooled by natural circulation, it is not adequate to use directly the same calculation methodologies from typical PWR and BWR.The low cooling flow rate and not having channels in the fuel elements (open-channel fuels) produce that most of the models and computer programs typically used must be carefully validated.As result of that process, an adequate calculation procedure for this reactor type was developed.In the present work, the thermal-hydraulic design criteria of the core and the design bases, the uncertainties factors, and the thermal margin results of the core are described.Despite that the methodology of DNBR calculation is under a validation process and considering the available calculation tools, it is possible to assure that the core fulfills the safety regulations in steady state conditions

  19. Economical evaluation of damaged vacuum insulation panels in buildings

    Science.gov (United States)

    Kim, Y. M.; Lee, H. Y.; Choi, G. S.; Kang, J. S.

    2015-12-01

    In Korea, thermal insulation standard of buildings have been tightened annually to satisfy the passive house standard from the year 2009. The current domestic policies about disseminating green buildings are progressively conducted. All buildings should be the zero energy building in the year 2025, obligatorily. The method is applied to one of the key technologies for high-performance insulation for zero energy building. The vacuum insulation panel is an excellent high performance insulation. But thermal performance of damaged vacuum insulation panels is reduced significantly. In this paper, the thermal performance of damaged vacuum insulation panels was compared and analyzed. The measurement result of thermal performance depends on the core material type. The insulation of building envelope is usually selected by economic feasibility. To evaluate the economic feasibility of VIPs, the operation cost was analyzed by simulation according to the types and damaged ratio of VIPs

  20. Assessment of core damage models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

    International Nuclear Information System (INIS)

    Coryell, E.W.

    1991-01-01

    The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel

  1. Subharmonic excitation in an HTGR core

    International Nuclear Information System (INIS)

    Bezler, P.; Curreri, J.R.

    1977-01-01

    The occurrence of subharmonic resonance in a series of blocks with clearance between blocks and with springs on the outer most ends is the subject of this paper. This represents an HTGR core response to an earthquake input. An analytical model of the cross section of this type of core is a series of blocks arranged horizontally between outer walls. Each block represents many graphite hexagonal core elements acting in unison as a single mass. The blocks are of unequal size to model the true mass distribution through the core. Core element elasticity and damping characteristics are modeled with linear spring and viscous damping units affixed to each block. The walls and base represent the core barell or core element containment structure. For forced response calculations, these boundaries are given prescribed motions. The clearance between each block could be the same or different with the total clearance duplicating that of the entire core. Spring packs installed between the first and last block and the boundaries model the boundary elasticity. The system non-linearity is due to the severe discontinuity in the interblock elastic forces when adjacent blocks collide. A computer program using a numerical integration scheme was developed to solve for the response of the system to arbitrary inputs

  2. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  3. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Moranchel y R, M.

    2012-01-01

    chemical of steel components concentration profiles after irradiated, among others. MCNPX version 2005 and SRIM-2011 there are the computer codes used in the Monte Carlo simulation for analyzing the vessel damage theoretically due to penetration and the radiation transport of the neutrons coming of the reactor core. With MCNPX was determined the profile of neutrons penetration, the photons flux, electrons, alphas, protons deuterons, etc, in the interior of the vessel, the profile of energy deposited by all particles generated in the interaction of neutrons with the steel atomic structures, the profile displacement, the regions of greater damage as well as the total damage over the life of the reactor, among many others. With SRIM-2011 was determined the atomic displacements profile of each atomic species in the steel (Fe, Ni, Mn, Cr, Mo, V, Cu, etc.), the energy deposited profile by each atomic species, the profile of damage in the interior of the vessel, the greater damage region as well as the total damage throughout the life of the nuclear reactor. The theoretical results have enabled understanding of physical phenomena occurring in the crystalline structure of steel, by influence of irradiation neutron to determine the region of greatest possible damage in the crock. The theoretical results were compared with experimental results and those of international references in this field of research. Among other results, is determined that the region of greater damage occurs in the first millimeter of depth of the vessel and confirms its full integrity after eight years of neutron irradiation and to not less than 60 years of life with a thickness of 12 cm. (Author)

  4. Measuring device for the coolant flowrate in a reactor core

    International Nuclear Information System (INIS)

    Sawa, Toshihiko.

    1983-01-01

    Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

  5. Overview of JSPS Core-to-Core Program: Forming Research and Educational Hubs of Medical Physics.

    Science.gov (United States)

    Koizumi, Masahiko; Takashina, Masaaki

    To foster medical physicists, we introduce the achievement we made since 2011 under the national research project of the Japan Society for the Promotion of Science (JSPS) Core-to-Core program; 'Forming Research and Educational Hubs of Medical Physics.' On this basis and under the JSPS program, we promoted research and educational exchange with Indiana University (IU) in USA, University of Groningen (The UG) in the Netherland and other cooperating institutions such as University of Minnesota (UM).A total of 23 students and researchers were sent. UG accepted the most among three institutions. In turn, 12 foreign researchers including post-doctor fellows came to Japan for academic seminars or educational lectures.Fifteen international seminars were held; 8 in Japan, 4 in USA, and 3 in the Netherland.Lots of achievement were made through these activities in 5 years. Total of 23 research topics at the international conferences were presented. Total of 12 articles were published in international journals.This program clearly promoted the establishment of international collaboration, and many young researchers and graduate students were exchanged and collaborated with foreign researchers.

  6. Mitochondrial iron accumulation exacerbates hepatic toxicity caused by hepatitis C virus core protein

    Energy Technology Data Exchange (ETDEWEB)

    Sekine, Shuichi; Ito, Konomi; Watanabe, Haruna; Nakano, Takafumi [Laboratory of Biopharmaceutics, Graduate School of Pharmaceutical Sciences, Chiba University, 1-8-1 Inohana, Chuo-ku, Chiba 260-8675 (Japan); Moriya, Kyoji; Shintani, Yoshizumi; Fujie, Hajime; Tsutsumi, Takeya; Miyoshi, Hideyuki; Fujinaga, Hidetake; Shinzawa, Seiko; Koike, Kazuhiko [Department of Internal Medicine, Graduate School of Medicine, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8655 (Japan); Horie, Toshiharu, E-mail: t.horie@thu.ac.jp [Laboratory of Biopharmaceutics, Graduate School of Pharmaceutical Sciences, Chiba University, 1-8-1 Inohana, Chuo-ku, Chiba 260-8675 (Japan)

    2015-02-01

    Patients with long-lasting hepatitis C virus (HCV) infection are at major risk of hepatocellular carcinoma (HCC). Iron accumulation in the livers of these patients is thought to exacerbate conditions of oxidative stress. Transgenic mice that express the HCV core protein develop HCC after the steatosis stage and produce an excess of hepatic reactive oxygen species (ROS). The overproduction of ROS in the liver is the net result of HCV core protein-induced dysfunction of the mitochondrial respiratory chain. This study examined the impact of ferric nitrilacetic acid (Fe-NTA)-mediated iron overload on mitochondrial damage and ROS production in HCV core protein-expressing HepG2 (human HCC) cells (Hep39b cells). A decrease in mitochondrial membrane potential and ROS production were observed following Fe-NTA treatment. After continuous exposure to Fe-NTA for six days, cell toxicity was observed in Hep39b cells, but not in mock (vector-transfected) HepG2 cells. Moreover, mitochondrial iron ({sup 59}Fe) uptake was increased in the livers of HCV core protein-expressing transgenic mice. This increase in mitochondrial iron uptake was inhibited by Ru360, a mitochondrial Ca{sup 2+} uniporter inhibitor. Furthermore, the Fe-NTA-induced augmentation of mitochondrial dysfunction, ROS production, and cell toxicity were also inhibited by Ru360 in Hep39b cells. Taken together, these results indicate that Ca{sup 2+} uniporter-mediated mitochondrial accumulation of iron exacerbates hepatocyte toxicity caused by the HCV core protein. - Highlights: • Iron accumulation in the livers of patients with hepatitis C virus (HCV) infection is thought to exacerbate oxidative stress. • The impact of iron overload on mitochondrial damage and ROS production in HCV core protein-expressing cells were examined. • Mitochondrial iron uptake was increased in the livers of HCV core protein-expressing transgenic mice. • Ca{sup 2+} uniporter-mediated mitochondrial accumulation of iron exacerbates

  7. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 7--12). Volume 2, Part 1B

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specific shutdown accidents would be useful

  8. An in-fiber integrated optofluidic device based on an optical fiber with an inner core.

    Science.gov (United States)

    Yang, Xinghua; Yuan, Tingting; Teng, Pingping; Kong, Depeng; Liu, Chunlan; Li, Entao; Zhao, Enming; Tong, Chengguo; Yuan, Libo

    2014-06-21

    A new kind of optofluidic in-fiber integrated device based on a specially designed hollow optical fiber with an inner core is designed. The inlets and outlets are built by etching the surface of the optical fiber without damaging the inner core. A reaction region between the end of the fiber and a solid point obtained after melting is constructed. By injecting samples into the fiber, the liquids can form steady microflows and react in the region. Simultaneously, the emission from the chemiluminescence reaction can be detected from the remote end of the optical fiber through evanescent field coupling. The concentration of ascorbic acid (AA or vitamin C, Vc) is determined by the emission intensity of the reaction of Vc, H2O2, luminol, and K3Fe(CN)6 in the optical fiber. A linear sensing range of 0.1-3.0 mmol L(-1) for Vc is obtained. The emission intensity can be determined within 2 s at a total flow rate of 150 μL min(-1). Significantly, this work presents information for the in-fiber integrated optofluidic devices without spatial optical coupling.

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

    Science.gov (United States)

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

    2018-01-30

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

  10. Core design aspects of SNR 2

    International Nuclear Information System (INIS)

    Wehmann, U.K.

    1987-01-01

    The paper describes in its first part the main characteristics of the core of the SNR 2 fast breeder reactor which is being planned within the European collaboration on fast breeder reactors. In the second part some core design aspects are discussed. The fuel element management with an inwards shuffling after each cycle is illustrated which offers advantages with respect to linear rating, steel damage and average discharge burnup. For this management, the full three-dimensional power and burnup history has been calculated and some typical results are presented. The shutdown requirements and the capabilities of the two shutdown systems of SNR 2 are discussed. The necessity for a reliable surveillance of the power distribution is demonstrated by the pronounced power tilts in case of the unintentional withdrawal of an absorber rod. Finally, a short review of the main nuclear design methods and their validation with help of the evaluation of experiments in zero power facilities and power reactors is given

  11. Application of Nonlinear Elastic Resonance Spectroscopy For Damage Detection In Concrete: An Interesting Story

    Energy Technology Data Exchange (ETDEWEB)

    Byers, Loren W. [Los Alamos National Laboratory; Ten Cate, James A. [Los Alamos National Laboratory; Johnson, Paul A. [Los Alamos National Laboratory

    2012-06-28

    Nonlinear resonance ultrasound spectroscopy experiments conducted on concrete cores, one chemically and mechanically damaged by alkali-silica reactivity, and one undamaged, show that this material displays highly nonlinear wave behavior, similar to many other damaged materials. They find that the damaged sample responds more nonlinearly, manifested by a larger resonant peak and modulus shift as a function of strain amplitude. The nonlinear response indicates that there is a hysteretic influence in the stress-strain equation of state. Further, as in some other materials, slow dynamics are present. The nonlinear response they observe in concrete is an extremely sensitive indicator of damage. Ultimately, nonlinear wave methods applied to concrete may be used to guide mixing, curing, or other production techniques, in order to develop materials with particular desired qualities such as enhanced strength or chemical resistance, and to be used for damage inspection.

  12. Few-Group Transport Analysis of the Core-Reflector Problem in Fast Reactor Cores via Equivalent Group Condensation and Local/Global Iteration

    International Nuclear Information System (INIS)

    Won, Jong Hyuck; Cho, Nam Zin

    2011-01-01

    In deterministic neutron transport methods, a process called fine-group to few-group condensation is used to reduce the computational burden. However, recent results on the core-reflector problem in fast reactor cores show that use of a small number of energy groups has limitation to describe neutron flux around core reflector interface. Therefore, researches are still ongoing to overcome this limitation. Recently, the authors proposed I) direct application of equivalently condensed angle-dependent total cross section to discrete ordinates method to overcome the limitation of conventional multi-group approximations, and II) local/global iteration framework in which fine-group discrete ordinates calculation is used in local problems while few-group transport calculation is used in the global problem iteratively. In this paper, an analysis of the core-reflector problem is performed in few-group structure using equivalent angle-dependent total cross section with local/global iteration. Numerical results are obtained under S 12 discrete ordinates-like transport method with scattering cross section up to P1 Legendre expansion

  13. On Monte Carlo estimation of radiation damage in light water reactor systems

    International Nuclear Information System (INIS)

    Read, Edward A.; Oliveira, Cassiano R.E. de

    2010-01-01

    There has been a growing need in recent years for the development of methodologies to calculate damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is discuss and highlight the main issues associated with the calculation of radiation damage factors utilizing the Monte Carlo method. Among these issues are: particle tracking and tallying in complex geometries, dpa calculation methodology, coupled fuel depletion and uncertainty propagation. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and burnup are assessed for radiation damage calculations and its capability demonstrated and compared to those of the MCNP code for dpa calculations of a typical LWR configuration involving the core vessel and the downcomer. (author)

  14. Degraded core studies at INEL

    International Nuclear Information System (INIS)

    Buescher, B.J.; Howe, T.M.; Miller, R.W.

    1982-01-01

    During 1980, planning of prototypical severe fuel damage tests to be conducted in the Power Burst Facility (PBF) to investigate fuel behavior in severe accidents up to temperatures of 2400 0 K was initiated. This first series of tests is designated Phase I. Also, a code development effort was initiated to provide a reliable predictive tool for core behavior during severe accidents. During 1981, an assessment of capabilities and preliminary planning were begun for an in-pile experimental program to investigate the behavior of larger arrays of previously irradiated fuel rods at temperatures through UO 2 melting. This latter series of tests is designated Phase II

  15. Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

    International Nuclear Information System (INIS)

    Petroski, Robert; Forget, Benoit; Forsberg, Charles

    2013-01-01

    Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

  16. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

  17. Candidate molten salt investigation for an accelerator driven subcritical core

    Science.gov (United States)

    Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

    2013-09-01

    We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

  18. Progressive Damage Modeling of Durable Bonded Joint Technology

    Science.gov (United States)

    Leone, Frank A.; Davila, Carlos G.; Lin, Shih-Yung; Smeltzer, Stan; Girolamo, Donato; Ghose, Sayata; Guzman, Juan C.; McCarville, Duglas A.

    2013-01-01

    The development of durable bonded joint technology for assembling composite structures for launch vehicles is being pursued for the U.S. Space Launch System. The present work is related to the development and application of progressive damage modeling techniques to bonded joint technology applicable to a wide range of sandwich structures for a Heavy Lift Launch Vehicle. The joint designs studied in this work include a conventional composite splice joint and a NASA-patented Durable Redundant Joint. Both designs involve a honeycomb sandwich with carbon/epoxy facesheets joined with adhesively bonded doublers. Progressive damage modeling allows for the prediction of the initiation and evolution of damage. For structures that include multiple materials, the number of potential failure mechanisms that must be considered increases the complexity of the analyses. Potential failure mechanisms include fiber fracture, matrix cracking, delamination, core crushing, adhesive failure, and their interactions. The joints were modeled using Abaqus parametric finite element models, in which damage was modeled with user-written subroutines. Each ply was meshed discretely, and layers of cohesive elements were used to account for delaminations and to model the adhesive layers. Good correlation with experimental results was achieved both in terms of load-displacement history and predicted failure mechanisms.

  19. Parameters affecting of Akkuyu’s safety assessment for severe core damages

    Directory of Open Access Journals (Sweden)

    Kavun Yusuf

    2015-01-01

    Full Text Available We have looked at all past core meltdowns (Three Mile Island, Chernobyl and Fukushima incidents and postulated the fourth one might be taking place in the future most probably in a newly built reactors anywhere of the earth in any type of NPP. The probability of this observation is high considering the nature of the machine and human interaction. Operation experience is a very significant parameter as well as the safety culture of the host nation. The concerns is not just a lack of experience with industry with the new comers, but also the infrastructure and established institutions who will be dealing with the Emergencies. Lack of trained and educated Emergency Response Organizations (ERO is a major concern. The culture on simple fire drills even makes the difference when a severe condition occurs in the industry. The study assumes the fourth event will be taking place at the Akkuyu NGS and works backwards as required by the “what went wrong ” scenarios and comes up with interesting results. The differences studied in depth to determine the impact to the severe accidents. The all four design have now core catchers. We have looked at the operator errors’like in TMI; Operator errors combined with design deficiencies(like in Chernobyl and natural disasters( like in Fukushima and found operator errors to be more probable event on the Akkuyu’s postulated next incident. With respect to experiences of the operators we do not have any data except for long and successful operating history of the Soviet design reactors up until the Chernobyl incident. Since the Akkuyu will be built, own and operated by the Russians we have found no alarming concerns at the moment. At the moment, there is no body be able to operate those units in Turkey. Turkey is planning to build the required manpower during the transition period. The resolution of the observed parameters lies to work and educate, train of the host nation and exercise together.

  20. Drilling history core hole DC-4

    International Nuclear Information System (INIS)

    1978-12-01

    Core hole DC-4 was completed at a depth of 3998 feet in December, 1978 by Boyles Brothers Drilling Company, Spokane, Washington, under subcontract to Fenix and Scission, Inc. The hole was cored for the US Department of Energy and the Rockwell Hanford Operations' Basalt Waste Isolation Program. Fenix and Sicsson, Inc. furnished the engineering, daily supervision of the cable tool and core drilling activities, and geological core logging for DC-4. Core hole DC-4 is located on the Hanford Site about 3 miles east of the Yakima Barricade and approximately 103 feet southwest of rotary hole DC-5, which was completed to 3990 feet in February, 1978. Hanford Site coordinates reported for hole DC-4 are north 49,385.62 feet and west 85,207.63 feet, and Washington State coordinates are north 454,468.73 feet and east 2,209,990.87 feet. No elevation survey is available for hole DC-4, but it is approximately 745 feet above mean sea level based upon the survey of hole DC-5, which has a reported elevation of 745.16 feet on the top of the 3-inch flange. The purpose of core hole DC-4 was to core drill vertically through the basalt and interbed units for stratigraphic depth determination and core collection, and to provide a borehole for hydrologic testing, cross-hole seismic shear, and pressure wave velocity studies with rotary hole DC-5. Hole DC-4 was drilled through the overburden into basalt bedrock by cable tool methods (0-623 feet) and continuously cored through the final interval (623 to 3998 feet).Core recovery was 95.8 percent of the total footage cored

  1. Radiation resistivity of pure-silica core image guide

    International Nuclear Information System (INIS)

    Hayami, H.; Ishitani, T.; Kishihara, O.; Suzuki, K.

    1988-01-01

    Radiation resistivity of pure-silica core image guides were investigated in terms of incremental spectral loss and quality of pictures transmitted through the image guides. Radiation-induced spectral losses were measured so as to clarify the dependences of radiation resistivity on such parameters as core materials (OH and Cl contents), picture element dimensions, (core packing density and cladding thickness), number of picture elements and drawing conditions. As the results, an image guide with OH-and Cl-free pure-silica core, 30-45% in core packing density, and 1.8 ∼ 2.2 μm in cladding thickness showed the lowest loss. The parameters to design this image guide were almost the same as those to obtain a image guide with good picture quality. Radiation resistivity of the image guide was not dependent on drawing conditions and number of picture elements, indicating that the image guide has large allowable in production conditions and that reliable quality is constantly obtained in production. Radiation resistivity under high total doses was evaluated using the image guide with the lowest radiation-induced loss. Maximum usable lengths of the image guide for practical use under specific high total doses and maximum allowable total doses for the image guide in specific lengths were extrapolated. Picture quality in terms of radiation-induced degradation in color fidelity in the pictures transmitted through image guides was quantitatively evaluated in the chromaticity diagram based on the CIE standard colorimetric system and in the color specification charts according to three attributes of colors. The image guide with the least spectral incremental loss gives the least radiation-induced degradation in color fidelity in the pictures as well. (author)

  2. Benefits of invasion prevention: Effect of time lags, spread rates, and damage persistence

    Science.gov (United States)

    Rebecca S. Epanchin-Niell; Andrew M. Liebhold

    2015-01-01

    Quantifying economic damages caused by invasive species is crucial for cost-benefit analyses of biosecurity measures. Most studies focus on short-term damage estimates, but evaluating exclusion or prevention measures requires estimates of total anticipated damages from the time of establishment onward. The magnitude of such damages critically depends on the timing of...

  3. Evaluation of fatigue damage for wind turbine blades using acoustic emission

    Energy Technology Data Exchange (ETDEWEB)

    Jee, Hyun Sup; Ju, No Hoe [Korea Institute of Materials Science, Changwon (Korea, Republic of); So, Cheal Ho [Dongshin University, Naju (Korea, Republic of); Lee, Jong Kyu [Dept. of Physics, Pukyung National University, Busan (Korea, Republic of)

    2015-06-15

    In this study, the flap fatigue test of a 48 m long wind turbine blade was performed for 1 million cycles to evaluate the characteristics of acoustic emission signals generated from fatigue damage of the wind blades. As the number of hits and total energy continued to increase during the first 0.6 million cycles, blade damage was constant. The rise-time result showed that the major aspects of damage were initiation and propagation of matrix cracks. In addition, the signal analysis of each channel showed that the most seriously damaged sections were the joint between the skin and spar, 20 m from the connection, and the spot of actual damage was observable by visual inspection. It turned out that the event source location was related to the change in each channel{sup s} total energy. It is expected that these findings will be useful for the optimal design of wind turbine blades.

  4. Analysis of the SPERT III E-core experiment using the EUREKA-2 code

    International Nuclear Information System (INIS)

    Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

    1986-09-01

    EUREKA-2, a coupled nuclear thermal hydrodynamic kinetic code, was adapted for the testing of models and methods. Code evaluations were made with the reactivity addition experiments of the SPERT III E-Core, a slightly enriched oxide core. The code was tested for non damaging power excursions including a wide range of initial operating conditions, such as cold-startup, hot-startup, hot-standby and operating-power initial conditions. Comparisons resulted in a good agreement within the experimental errors between calculated and experimental power, energy, reactivity and clad surface temperature. (author)

  5. Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

  6. Lymphocyte DNA damage and oxidative stress in patients with iron deficiency anemia.

    Science.gov (United States)

    Aslan, Mehmet; Horoz, Mehmet; Kocyigit, Abdurrahim; Ozgonül, Saadet; Celik, Hakim; Celik, Metin; Erel, Ozcan

    2006-10-10

    Oxidant stress has been shown to play an important role in the pathogenesis of iron deficiency anemia. The aim of this study was to investigate the association between lymphocyte DNA damage, total antioxidant capacity and the degree of anemia in patients with iron deficiency anemia. Twenty-two female with iron deficiency anemia and 22 healthy females were enrolled in the study. Peripheral DNA damage was assessed using alkaline comet assay and plasma total antioxidant capacity was determined using an automated measurement method. Lymphocyte DNA damage of patients with iron deficiency anemia was significantly higher than controls (ptotal antioxidant capacity was significantly lower (ptotal antioxidant capacity and hemoglobin levels (r=0.706, ptotal antioxidant capacity and hemoglobin levels were negatively correlated with DNA damage (r=-0.330, p<0.05 and r=-0.323, p<0.05, respectively). In conclusion, both oxidative stress and DNA damage are increased in IDA patients. Increased oxidative stress seems as an important factor that inducing DNA damage in those IDA patients. The relationships of oxidative stress and DNA damage with the severity of anemia suggest that both oxidative stress and DNA damage may, in part, have a role in the pathogenesis of IDA.

  7. Ubiquitin ligase activity of TFIIH and the transcriptional response to DNA damage.

    Science.gov (United States)

    Takagi, Yuichiro; Masuda, Claudio A; Chang, Wei-Hau; Komori, Hirofumi; Wang, Dong; Hunter, Tony; Joazeiro, Claudio A P; Kornberg, Roger D

    2005-04-15

    Core transcription factor (TF) IIH purified from yeast possesses an E3 ubiquitin (Ub) ligase activity, which resides, at least in part, in a RING finger (RNF) domain of the Ssl1 subunit. Yeast strains mutated in the Ssl1 RNF domain are sensitive to ultraviolet (UV) light and to methyl methanesulfonate (MMS). This increased sensitivity to DNA-damaging agents does not reflect a deficiency in nucleotide excision repair. Rather, it correlates with reduced transcriptional induction of genes involved in DNA repair, suggesting that the E3 Ub ligase activity of TFIIH mediates the transcriptional response to DNA damage.

  8. Study on in-core fuel management for CNP1500 nuclear power plant

    International Nuclear Information System (INIS)

    Li Dongsheng

    2005-10-01

    CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

  9. CONSIDER - Core Outcome Set in IAD Research: study protocol for establishing a core set of outcomes and measurements in incontinence-associated dermatitis research.

    Science.gov (United States)

    Van den Bussche, Karen; De Meyer, Dorien; Van Damme, Nele; Kottner, Jan; Beeckman, Dimitri

    2017-10-01

    This study protocol describes the methodology for the development of a core set of outcomes and a core set of measurements for incontinence-associated dermatitis. Incontinence is a widespread disorder with an important impact on quality of life. One of the most common complications is incontinence-associated dermatitis, resulting from chemical and physical irritation of the skin barrier, triggering inflammation and skin damage. Managing incontinence-associated dermatitis is an important challenge for nurses. Several interventions have been assessed in clinical trials, but heterogeneity in study outcomes complicates the comparability and standardization. To overcome this challenge, the development of a core outcome set, a minimum set of outcomes and measurements to be assessed in clinical research, is needed. A project team, International Steering Committee and panelists will be involved to guide the development of the core outcome set. The framework of the Harmonizing Outcomes Measures for Eczema roadmap endorsed by Cochrane Skin Group Core Outcomes Set Initiative, is used to inform the project design. A systematic literature review, interviews to integrate the patients' perspective and a consensus study with healthcare researchers and providers using the Delphi procedure will be performed. The project was approved by the Ethics review Committee (April 2016). This is the first project that will identify a core outcome set of outcomes and measurements for incontinence-associated dermatitis research. A core outcome set will reduce possible reporting bias, allow results comparisons and statistical pooling across trials and strengthen evidence-based practice and decision-making. This project has been registered in the Core Outcome Measures in Effectiveness Trials (COMET) database and is part of the Cochrane Skin Group Core Outcomes Set Initiative (CSG-COUSIN). © 2016 John Wiley & Sons Ltd.

  10. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices E (Sections E.1--E.8). Volume 2, Part 3A

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

  11. A concise synthesis of the cortistatin core

    OpenAIRE

    Dai, Mingji; Danishefsky, Samuel J.

    2008-01-01

    We describe a concise and convergent route to the core matrix of the cortistatin steroidal alkaloids. The salient features of the synthesis are the Snieckus cascade methodology and the Masamune alkylative dearomatization. This chemistry lends itself to a total synthesis of the cortistatins and to the development of a SAR program based on diverted total synthesis.

  12. Results and Prospects of Development of Works on Structural Core Materials for Russian Fast Reactors

    International Nuclear Information System (INIS)

    Nikitina, A.A.; Ageev, V.S.; Leontyeva-Smirnova, M.V.; Mitrofanova, N.M.; Tselishchev, A.V.

    2015-01-01

    The strategy of development of atomic energy in Russia in the first half of XXI century contemplates construction and putting in operation of fast reactors of new generation with different types of coolant: sodium (BN-800, BN-1200, MBIR), lead (BREST-OD-300) and lead-bismuth eutectic (SVBR-100). For assurance of the working capacity of reactors that are under construction and achievement of economically reasonable burn-up of nuclear fuel the structural core materials with necessary level of radiation resistance, heat resistance, corrosion resistance to products of fuel fission, corrosion resistance in coolant and in water must be developed and justified. For sodium cooled reactors the key challenge is creation of radiation resistant and heat resistant cladding materials, which must ensure the achievement of damage doses at least 140 dpa. The solution of this problem is provided by phased use as cladding materials of austenitic steels ChS68 and EK164 (maximum damage doses ~ 92 and ~110-115 dpa, respectively), precipitation-hardening heat resistant ferritic-martensitic steels EK181 and ChS139 (maximum damage dose ~140 dpa) and oxide dispersion strengthened (ODS) steels (maximum damage dose more than 140 dpa). For development of core materials for reactors with lead and lead-bismuth eutectic coolants the most serious challenge is corrosion resistance of materials in coolant. Therefore at present time a very wide range of works on study of corrosion resistance of candidate materials is carrying out. As the basic material for the cladding tubes is considered a ferritic-martensitic steel EP823 with high silicon content. In this report the main results of works on justification of the working capacity of materials of different classes in respect to use it in cores of operating and prospective fast reactors with different types of coolant and prospects of further development of works are presented. (author)

  13. Calculation of ex-core detector responses

    Energy Technology Data Exchange (ETDEWEB)

    Wouters, R. de; Haedens, M. [Tractebel Engineering, Brussels (Belgium); Baenst, H. de [Electrabel, Brussels (Belgium)

    2005-07-01

    The purpose of this work carried out by Tractebel Engineering, is to develop and validate a method for predicting the ex-core detector responses in the NPPs operated by Electrabel. Practical applications are: prediction of ex-core calibration coefficients for startup power ascension, replacement of xenon transients by theoretical predictions, and analysis of a Rod Drop Accident. The neutron diffusion program PANTHER calculates node-integrated fission sources which are combined with nodal importance representing the contribution of a neutron born in that node to the ex-core response. These importance are computed with the Monte Carlo program MCBEND in adjoint mode, with a model of the whole core at full power. Other core conditions are treated using sensitivities of the ex-core responses to water densities, computed with forward Monte Carlo. The Scaling Factors (SF), or ratios of the measured currents to the calculated response, have been established on a total of 550 in-core flux maps taken in four NPPs. The method has been applied to 15 startup transients, using the average SF obtained from previous cycles, and to 28 xenon transients, using the SF obtained from the in-core map immediately preceding the transient. The values of power (P) and axial offset (AOi) reconstructed with the theoretical calibration agree well with the measured values. The ex-core responses calculated during a rod drop transient have been successfully compared with available measurements, and with theoretical data obtained by alternative methods. In conclusion, the method is adequate for the practical applications previously listed. (authors)

  14. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  15. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 1--6). Volume 2, Part 1A

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1992-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown written specifically for shutdown accidents would be useful. This document presents Chapters 1--6 of the report

  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. Economic measurement of environment damages

    Energy Technology Data Exchange (ETDEWEB)

    Krawiec, F.

    1980-05-01

    The densities, energy consumption, and economic development of the increasing population exacerbate environmental degradation. Air and water pollution is a major environmental problem affecting life and health, outdoor recreation, household soiling, vegetation, materials, and production. The literature review indicated that numerous studies have assessed the physical and monetary damage to populations at risk from excessive concentrations of major air and water pollutants-sulfur dioxide, total suspended particulate matter, oxidants, and carbon monoxide in air; and nutrients, oil, pesticides, and toxic metals and others in water. The measurement of the damages was one of the most controversial issues in pollution abatement. The methods that have been used to estimate the societal value of pollution abatement are: (1) chain of effects, (2) market approaches, and (3) surveys. National gross damages of air pollution of $20.2 billion and of water pollution of $11.1 billion for 1973 are substantial. These best estimates, updated for the economic and demographic conditions, could provide acceptable control totals for estimating and predicting benefits and costs of abating air and water pollution emissions. The major issues to be resolved are: (1) lack of available noneconomic data, (2) theoretical and empirical difficulties of placing a value on human life and health and on benefits such as aesthetics, and (3) lack of available demographic and economic data.

  18. Uncertainty in urban flood damage assessment due to urban drainage modelling and depth-damage curve estimation.

    Science.gov (United States)

    Freni, G; La Loggia, G; Notaro, V

    2010-01-01

    because a large part of the total uncertainty is dependent on depth-damage curves. Improving the estimation of these curves may provide better results in term of uncertainty reduction than the adoption of detailed hydraulic models.

  19. SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

    International Nuclear Information System (INIS)

    Wilson, D.R.; Dutt, D.S.

    1979-05-01

    SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

  20. 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

  1. Core reset system design for linear induction accelerator

    International Nuclear Information System (INIS)

    Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

    2006-01-01

    A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

  2. Environmental damage costs in Iran by the energy sector

    International Nuclear Information System (INIS)

    Shafie-Pour, Majid; Ardestani, Mojtaba

    2007-01-01

    On the basis of the energy supply and demand, this paper assesses the environmental damage from air pollution in Iran using the Extern-E study that has extended over 10 years and is still in progress in the European Union (EU) commission. Damage costs were transferred from Western European practice to the conditions of Iran by scaling according to GDP per capital measured in PPP terms. Using this approach, the total health damage from air pollution in 2001 is assessed at about $7 billion; equivalent to 8.4% of nominal GDP. In the absence of price reform and control policies, it is estimated that damage in Iran will grow to $9 billion by 2019, in the money of 2001. This is equivalent to 10.9% of nominal GDP, i.e. a larger percentage of a larger GDP. Of this total, $8.4 billion comes from the transport sector. The damage cost to the global environment from the flaring of natural gas, assessed on the basis of a carbon price of $10/ton CO 2 and found to be approximately $600 million per year. This is equal to a little less than 1% of current GDP. There are larger costs associated with recovery and use of such gas, but equally there are large potential benefits

  3. Testosterone Depletion by Castration May Protect Mice from Heat-Induced Multiple Organ Damage and Lethality

    Directory of Open Access Journals (Sweden)

    Ruei-Tang Cheng

    2010-01-01

    Full Text Available When the vehicle-treated, sham-operated mice underwent heat stress, the fraction survival and core temperature at +4 h of body heating were found to be 5 of 15 and 34.4∘C±0.3∘C, respectively. Castration 2 weeks before the start of heat stress decreased the plasma levels of testosterone almost to zero, protected the mice from heat-induced death (fraction survival, 13/15 and reduced the hypothermia (core temperature, 37.3∘C. The beneficial effects of castration in ameliorating lethality and hypothermia can be significantly reduced by testosterone replacement. Heat-induced apoptosis, as indicated by terminal deoxynucleotidyl- transferase- mediatedαUDP-biotin nick end-labeling staining, were significantly prevented by castration. In addition, heat-induced neuronal damage, as indicated by cell shrinkage and pyknosis of nucleus, to the hypothalamus was also castration-prevented. Again, the beneficial effects of castration in reducing neuronal damage to the hypothalamus as well as apoptosis in multiple organs during heatstroke, were significantly reversed by testosterone replacement. The data indicate that testosterone depletion by castration may protect mice from heatstroke-induced multiple organ damage and lethality.

  4. Pollutant plume delineation from tree core sampling using standardized ranks

    DEFF Research Database (Denmark)

    Wahyudi, Agung; Bogaert, Patrick; Trapp, Stefan

    2012-01-01

    There are currently contradicting results in the literature about the way chloroethene (CE) concentrations from tree core sampling correlate with those from groundwater measurements. This paper addresses this issue by focusing on groundwater and tree core datasets in CE contaminated site, Czech...... Republic. Preliminary analyses revealed strongly and positively skewed distributions for the tree core dataset, with an intra-tree variability accounting for more than 80% of the total variability, while the spatial analyses based on variograms indicated no obvious spatial pattern for CE concentration...... groundwater and tree core measurements. Nonetheless, tree core sampling and analysis proved to be a quick and inexpensive semi-quantitative method and a useful tool....

  5. Fatigue-Damage Estimation and Control for Wind Turbines

    DEFF Research Database (Denmark)

    Barradas Berglind, Jose de Jesus

    How can fatigue-damage for control of wind turbines be represented? Fatigue-damage is indeed a crucial factor in structures such as wind turbines that are exposed to turbulent and rapidly changing wind flow conditions. This is relevant both in their design stage and during the control......, the inclusion of fatigue-damage within feedback control loops is of special interest. Four strategies in total are proposed in this work: three for the wind turbine level and one for the wind farm level. On one hand, the three strategies in the turbine level are based on hysteresis operators and strive......-damage estimation in wind turbine components, to the mixed objective operation of wind energy conversion systems, and to the synthesis of control strategies that include hysteresis operators....

  6. Evaluation report on CCTF core-II reflood test C2-6 (Run 64)

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu.

    1985-03-01

    In order to evaluate the effect of the radial power profile on the system behavior and the core thermal hydraulic behavior during the reflood phase of a PWR LOCA, a test was performed using the Cylindrical Core Test Facility(CCTF) with the flat radial power profile. The test was conducted with the same total core power as that of the steep radial power test C2-5(Run 63). Through the comparisons of the results from these two tests, the following conclusions were obtained: (1) The radial power profile in the core has weak effect on the thermal hydraulic behavior in the primary system except the core. (2) Almost the same differential pressure was observed at various elevations in the periphery of the core regardless of different radial power profile. The result suggests that the core differential pressure is determined mainly by the total power and the total stored energy rather than by the local power and the local stored energy. (3) The test results support the single channel core model with the average power rod used in the reactor safety analysis codes such as REFLA-1DS, WREM for the evaluation of the overall system behavior. (4) In the steep radial power test, the heat transfer coefficient in the central(high power) region was higher than that in the peripheral(low power) region. The tendency was not explained by the estimation with the heat transfer correlation developed by Murao and Sugimoto assuming that the void fraction was uniform in a horizontal cross section. It is necessary to study more the dependency of core heat transfer on the radial power profile in the wide core. (author)

  7. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  8. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  9. Analysis of Ricefield Land Damage in Denpasar City, Bali, Indonesia

    Science.gov (United States)

    Suyarto, R.; Wiyanti; Dibia, I. N.

    2018-02-01

    Soil as a natural resource, living area, environmental media, and factors of production including biomass production that supports human life and other living beings must be preserved, on the other hand, uncontrolled biomass production activities can cause soil damage, ultimately can threaten the survival of humans and other living things. Therefore, in order to control soil damage, first must inventories the soil condition data and its damage which then visualised in soil damage potential and soil damage status. The activities of the study are the preparation of a map of the initial soil conditions and the delineation of potentially land degradation distribution. Mapping results are used as work maps for verification on the field to take soil samples and create soil damage status. In general, Denpasar City have soil damage potential at very low, low until medium rate. Soil damage status in Denpasar City generally is low damage of bulk volume, total porosity, soil permeability and electrolyte conductivity which beyond limitation thresholds.

  10. Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

    1985-07-01

    Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

  11. Radiation damage in barium fluoride detector materials

    International Nuclear Information System (INIS)

    Levey, P.W.; Kierstead, J.A.; Woody, C.L.

    1988-01-01

    To develop radiation hard detectors, particularly for high energy physics studies, radiation damage is being studied in BaF 2 , both undoped and doped with La, Ce, Nd, Eu, Gd and Tm. Some dopants reduce radiation damage. In La doped BaF 2 they reduce the unwanted long lifetime luminescence which interferes with the short-lived fluorescence used to detect particles. Radiation induced coloring is being studied with facilities for making optical measurements before, during and after irradiation with 60 C0 gamma rays. Doses of 10 6 rad, or less, create only ionization induced charge transfer effects since lattice atom displacement damage is negligible at these doses. All crystals studied exhibit color center formation, between approximately 200 and 800 nm, during irradiation and color center decay after irradiation. Thus only measurements made during irradiation show the total absorption present in a radiation field. Both undoped and La doped BaF 2 develop damage at minimum detectable levels in the UV---which is important for particle detectors. For particle detector applications these studies must be extended to high dose irradiations with particles energetic enough to cause lattice atom displacement damage. In principle, the reduction in damage provided by dopants could apply to other applications requiring radiation damage resistant materials

  12. In Vivo Wear Performance of Cobalt-Chromium Versus Oxidized Zirconium Femoral Total Knee Replacements.

    Science.gov (United States)

    Gascoyne, Trevor C; Teeter, Matthew G; Guenther, Leah E; Burnell, Colin D; Bohm, Eric R; Naudie, Douglas R

    2016-01-01

    This study examines the damage and wear on the polyethylene (PE) inserts from 52 retrieved Genesis II total knee replacements to identify differences in tribological performance between matched pairs of cobalt-chromium (CoCr) and oxidized zirconium (OxZr) femoral components. Observer damage scoring and microcomputed tomography were used to quantify PE damage and wear, respectively. No significant differences were found between CoCr and OxZr groups in terms of PE insert damage, surface penetration, or wear. No severe damage such as cracking or delamination was noted on any of the 52 PE inserts. Observer damage scoring did not correlate with penetrative or volumetric PE wear. The more costly OxZr femoral component does not demonstrate clear tribological benefit over the standard CoCr component in the short term with this total knee replacement design. Copyright © 2016 Elsevier Inc. All rights reserved.

  13. Endogenous melatonin and oxidatively damaged guanine in DNA

    Directory of Open Access Journals (Sweden)

    Poulsen Henrik E

    2009-10-01

    Full Text Available Abstract Background A significant body of literature indicates that melatonin, a hormone primarily produced nocturnally by the pineal gland, is an important scavenger of hydroxyl radicals and other reactive oxygen species. Melatonin may also lower the rate of DNA base damage resulting from hydroxyl radical attack and increase the rate of repair of that damage. This paper reports the results of a study relating the level of overnight melatonin production to the overnight excretion of the two primary urinary metabolites of the repair of oxidatively damaged guanine in DNA. Methods Mother-father-daughter(s families (n = 55 were recruited and provided complete overnight urine samples. Total overnight creatinine-adjusted 6-sulphatoxymelatonin (aMT6s/Cr has been shown to be highly correlated with total overnight melatonin production. Urinary 8-oxo-7,8-dihydro-guanine (8-oxoGua results from the repair of DNA or RNA guanine via the nucleobase excision repair pathway, while urinary 8-oxo-7,8-dihydro-2'-deoxyguanosine (8-oxodG may possibly result from the repair of DNA guanine via the nucleotide excision repair pathway. Total overnight urinary levels of 8-oxodG and 8-oxoGua are therefore a measure of total overnight guanine DNA damage. 8-oxodG and 8-oxoGua were measured using a high-performance liquid chromatography-electrospray ionization tandem mass spectrometry assay. The mother, father, and oldest sampled daughter were used for these analyses. Comparisons between the mothers, fathers, and daughters were calculated for aMT6s/Cr, 8-oxodG, and 8-oxoGua. Regression analyses of 8-oxodG and 8-oxoGua on aMT6s/Cr were conducted for mothers, fathers, and daughters separately, adjusting for age and BMI (or weight. Results Among the mothers, age range 42-80, lower melatonin production (as measured by aMT6s/CR was associated with significantly higher levels of 8-oxodG (p Conclusion Low levels of endogenous melatonin production among older individuals may lead to

  14. Detection of Ballast Damage by In-Situ Vibration Measurement of Sleepers

    Science.gov (United States)

    Lam, H. F.; Wong, M. T.; Keefe, R. M.

    2010-05-01

    Ballasted track is one of the most important elements of railway transportation systems worldwide. Owing to its importance in railway safety, many monitoring and evaluation methods have been developed. Current railway track monitoring systems are comprehensive, fast and efficient in testing railway track level and alignment, rail gauge, rail corrugation, etc. However, the monitoring of ballast condition still relies very much on visual inspection and core tests. Although extensive research has been carried out in the development of non-destructive methods for ballast condition evaluation, a commonly accepted and cost-effective method is still in demand. In Hong Kong practice, if abnormal train vibration is reported by the train operator or passengers, permanent way inspectors will locate the problem area by track geometry measurement. It must be pointed out that visual inspection can only identify ballast damage on the track surface, the track geometry deficiencies and rail twists can be detected using a track gauge. Ballast damage under the sleeper loading area and the ballast shoulder, which are the main factors affecting track stability and ride quality, are extremely difficult if not impossible to be detected by visual inspection. Core test is a destructive test, which is expensive, time consuming and may be disruptive to traffic. A fast real-time ballast damage detection method that can be implemented by permanent way inspectors with simple equipment can certainly provide valuable information for engineers in assessing the safety and riding quality of ballasted track systems. The main objective of this paper is to study the feasibility in using the vibration characteristics of sleepers in quantifying the ballast condition under the sleepers, and so as to explore the possibility in developing a handy method for the detection of ballast damage based on the measured vibration of sleepers.

  15. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Jang, J. S.; Kim, D. W.

    2002-03-01

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  16. Probabilistic risk assessment (PRA) on the effectiveness of a core rescue system (SSN) for PWRs

    International Nuclear Information System (INIS)

    Petrangeli, G.; Valeri, A.

    1983-01-01

    Safety systems for the prevention of LWR core severe damage have recently been studied, which are based on automatic primary system depressurization and on borated water injection by low pressure accumulators. These systems have been named Core Rescue System (SSN). The present study evaluates the reduction in core melt probability brought about by the installation of a SSN system on the RSS (WASH 1400) PWR plant (Surry 1). The calculated result is a core melt probability reduction factor of about 250. Taking into account the possible effect of external or internal unknown events of negligible, yet undefined, probability it is concluded that a SSN system can make a plant ten times safer. The first part of a review report by Prof. N.C.Rasmussen, MIT, dealing with general comment, is attached

  17. Fracture Behaviours in Compression-loaded Triangular Corrugated Core Sandwich Panels

    Directory of Open Access Journals (Sweden)

    Zaid N.Z.M.

    2016-01-01

    Full Text Available The failure modes occurring in sandwich panels based on the corrugations of aluminium alloy, carbon fibre-reinforced plastic (CFRP and glass fibre-reinforced plastic (GFRP are analysed in this work. The fracture behaviour of these sandwich panels under compressive stresses is determined through a series of uniform lateral compression performed on samples with different cell wall thicknesses. Compression test on the corrugated-core sandwich panels were conducted using an Instron series 4505 testing machine. The post-failure examinations of the corrugated-core in different cell wall thickness were conducted using optical microscope. Load-displacement graphs of aluminium alloy, GFRP and CFRP specimens were plotted to show progressive damage development with five unit cells. Four modes of failure were described in the results: buckling, hinges, delamination and debonding. Each of these failure modes may dominate under different cell wall thickness or loading condition, and they may act in combination. The results indicate that thicker composites corrugated-core panels tend can recover more stress and retain more stiffness. This analysis provides a valuable insight into the mechanical behaviour of corrugated-core sandwich panels for use in lightweight engineering applications.

  18. Candidate molten salt investigation for an accelerator driven subcritical core

    International Nuclear Information System (INIS)

    Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

    2013-01-01

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated

  19. Candidate molten salt investigation for an accelerator driven subcritical core

    Energy Technology Data Exchange (ETDEWEB)

    Sooby, E., E-mail: soobyes@tamu.edu [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Baty, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Beneš, O. [European Commission, DG Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); McIntyre, P.; Pogue, N. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Salanne, M. [Université Pierre et Marie Curie, CNRS, Laboratoire PECSA, F-75005 Paris (France); Sattarov, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States)

    2013-09-15

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated.

  20. Radiation-induced liver damage

    International Nuclear Information System (INIS)

    Marcial, V.A.; Santiago-Delpin, E.A.; Lanaro, A.E.; Castro-Vita, H.; Arroyo, G.; Moscol, J.A.; Gomez, C.; Velazquez, J.; Prado, K.

    1977-01-01

    Due to the recent increase in the use of radiation therapy in the treatment of cancer with or without chemotherapy, the risk of liver radiation damage has become a significant concern for the radiotherapist when the treated tumour is located in the upper abdomen or lower thorax. Clinically evident radiation liver damage may result in significant mortality, but at times patients recover without sequelae. The dose of 3000 rads in 3 weeks to the entire liver with 5 fractions per week of 200 rads each, seems to be tolerated well clinically by adult humans. Lower doses may lead to damage when used in children, when chemotherapy is added, as in recent hepatectomy cases, and in the presence of pre-existent liver damage. Reduced fractionation may lead to increased damage. Increased fractionation, limitation of the dose delivered to the entire liver, and restriction of the high dose irradiation volume may afford protection. With the aim of studying the problems of hepatic radiation injury in humans, a project of liver irradiation in the dog is being conducted. Mongrel dogs are being conditioned, submitted to pre-irradiation studies (haemogram, blood chemistry, liver scan and biopsy), irradiated under conditions resembling human cancer therapy, and submitted to post-irradiation evaluation of the liver. Twenty-two dogs have been entered in the study but only four qualify for the evaluation of all the study parameters. It has been found that dogs are susceptible to liver irradiation damage similar to humans. The initial mortality has been high mainly due to non-radiation factors which are being kept under control at the present phase of the study. After the initial experiences, the study will involve variations in total dose and fractionation, and the addition of anticoagulant therapy for possible prevention of radiation liver injury. (author)

  1. 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)

  2. Magnetic properties of Ni/Au core/shell studied by Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Masrour, R., E-mail: rachidmasrour@hotmail.com [Laboratory of Materials, Processes, Environment and Quality, Cady Ayyed University, National School of Applied Sciences, Sidi Bouzid, Safi, 63 4600 (Morocco); LMPHE (URAC 12), Faculté des Sciences, Université Mohammed V-Agdal, Av. Ibn Batouta, B.P. 1014, Rabat (Morocco); Bahmad, L. [LMPHE (URAC 12), Faculté des Sciences, Université Mohammed V-Agdal, Av. Ibn Batouta, B.P. 1014, Rabat (Morocco); Hamedoun, M. [Institute of Nanomaterials and Nanotechnologies, MAScIR, Rabat (Morocco); Benyoussef, A. [LMPHE (URAC 12), Faculté des Sciences, Université Mohammed V-Agdal, Av. Ibn Batouta, B.P. 1014, Rabat (Morocco); Institute of Nanomaterials and Nanotechnologies, MAScIR, Rabat (Morocco); Hassan II Academy of Science and Technology, Rabat (Morocco); Hlil, E.K. [Institut Néel, CNRS et Université Joseph Fourier, BP 166, F-38042 Grenoble cedex 9 (France)

    2014-01-10

    The magnetic properties of ferromagnetic Ni/Au core/shell have been studied using Monte Carlo simulations within the Ising model framework. The considered Hamiltonian includes the exchange interactions between Ni–Ni, Au–Au and Ni–Au and the external magnetic field. The thermal total magnetizations and total magnetic susceptibilities of core/shell Ni/Au are computed. The critical temperature is deduced. The exchange interaction between Ni and Au atoms is obtained. In addition, the total magnetizations versus the external magnetic field and crystal filed for different temperature are also established.

  3. Total thyroidectomy as primary elective procedure in multinodular thyroid disease

    International Nuclear Information System (INIS)

    Sheikh, I.A.; Haider, I.Z.; Haroon, A.; Ashfaq, M.

    2009-01-01

    Multinodular goitre is one of the commonest thyroid diseases encountered in the practice of surgery. The most common surgery being performed for multinodular goitre is subtotal thyroidectomy. Total thyroidectomy is designed to remove all of the thyroid tissue. The objective of this study was to evaluate total thyroidectomy as a primary elective procedure for treatment of multinodular thyroid disease. This descriptive study was carried out at Combined Military Hospital Rawalpindi from June 2003 to September 2006. 88 patients of multinodular thyroid disease were included. Patients having evidence of recurrent laryngeal nerve damage, recurrent goitre, evidence of altered parathyroid functions or evidence of malignancy were excluded. All patients underwent total thyroidectomy by the same team of surgeons and the patients were closely followed up for postoperative complications especially in terms of recurrent laryngeal nerve damage and hypocalcaemic tetany. No major postoperative complication was noted. Only 1 patient (1.14%) developed unilateral recurrent laryngeal nerve damage and 2 patients (2.27%) developed transient hypocalcaemia that recovered quickly. Total thyroidectomy as a primary elective procedure in multinodular thyroid disease is a safe option and it removes the disease process completely, lowers local recurrence rates and avoids the substantial risks of re operative surgery. (author)

  4. Neutron Fluence, Dosimetry and Damage Response Determination in In-Core/Ex-Core Components of the VENUS CEN/SCK LWR Using 3-D Monte Carlo Simulations: NEA's VENUS-3 Benchmark

    International Nuclear Information System (INIS)

    Perlado, J. Manuel; Marian, Jaime; Sanz, Jesus Garcia

    2000-01-01

    Validating state-of-the-art methods used to predict fluence exposure to reactor pressure vessels (RPVs) has become an important issue in identifying the sources of uncertainty in the estimated RPV fluence for pressurized water reactors. This is a very important aspect in evaluating irradiation damage leading to the hardening and embrittlement of such structural components. One of the major benchmark experiments carried out to test three-dimensional methodologies is the VENUS-3 Benchmark Experiment in which three-dimensional Monte Carlo and S n codes have proved more efficient than synthesis methods. At the Instituto de Fusion Nuclear (DENIM) at the Universidad Politecnica de Madrid, a detailed full three-dimensional model of the Venus Critical Facility has been developed making use of the Monte Carlo transport code MCNP4B. The problem geometry and source modeling are described, and results, including calculated versus experimental (C/E) ratios as well as additional studies, are presented. Evidence was found that the great majority of C/E values fell within the 10% tolerance and most within 5%. Tolerance limits are discussed on the basis of evaluated data library and fission spectra sensitivity, where a value ranging between 10 to 15% should be accepted. Also, a calculation of the atomic displacement rate has been carried out in various locations throughout the reactor, finding that values of 0.0001 displacements per atom in external components such as the core barrel are representative of this type of reactor during a 30-yr time span

  5. Damage clustering in metals: Importance, advances and challenges

    International Nuclear Information System (INIS)

    Nordlund, K.; Sand, A.E.; Granberg, F.; Levo, E.; Djurabekova, F.

    2016-01-01

    The damage produced in metals has traditionally been primarily characterized in terms of the total damage production, which typically is first estimated with the dpa number. As discussed in previous meetings of this CRP, the dpa is not actually very well suited for typical dense metals, since the number it gives is typically about 3 times larger than the number of actual defects produced, and 30 times smaller than the actual number of defects produced. Hence we developed the improved arc-dpa and rpa standards, that give in a simple analytical form a defect number that does correspond well to MD and experimental data. Section 2 summarizes the development of the arc-dpa and rpa standards. In sections 3 and 4 we discuss the role of damage clustering in damage production

  6. Neutron and gamma irradiation damage to organic materials.

    Energy Technology Data Exchange (ETDEWEB)

    White, Gregory Von, II; Bernstein, Robert

    2012-04-01

    This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

  7. Size-exclusion chromatography using core-shell particles.

    Science.gov (United States)

    Pirok, Bob W J; Breuer, Pascal; Hoppe, Serafine J M; Chitty, Mike; Welch, Emmet; Farkas, Tivadar; van der Wal, Sjoerd; Peters, Ron; Schoenmakers, Peter J

    2017-02-24

    Size-exclusion chromatography (SEC) is an indispensable technique for the separation of high-molecular-weight analytes and for determining molar-mass distributions. The potential application of SEC as second-dimension separation in comprehensive two-dimensional liquid chromatography demands very short analysis times. Liquid chromatography benefits from the advent of highly efficient core-shell packing materials, but because of the reduced total pore volume these materials have so far not been explored in SEC. The feasibility of using core-shell particles in SEC has been investigated and contemporary core-shell materials were compared with conventional packing materials for SEC. Columns packed with very small core-shell particles showed excellent resolution in specific molar-mass ranges, depending on the pore size. The analysis times were about an order of magnitude shorter than what could be achieved using conventional SEC columns. Copyright © 2016 Elsevier B.V. All rights reserved.

  8. Heat, hydrogen peroxide, and UV resistance of Bacillus subtilis spores with increased core water content and with or without major DNA-binding proteins

    International Nuclear Information System (INIS)

    Popham, D.L.; Sengupta, S.; Setlow, P.

    1995-01-01

    Spores of a Bacillus subtilis strain with an insertion mutation in the dacB gene, which codes for an enzyme involved in spore cortex biosynthesis, have a higher core water content than wild-type spores. Spores lacking the two major α/β-type small, acid-soluble proteins (SASP) (termed a α - β - spores) have the same core water content as do wild-type spores, but α - β - dacB spores had more core water than did dacB spores. The resistance of α - β - , α - β - dacB, dacB, and wild-type spores to dry and moist heat, hydrogen peroxide, and UV radiation has been determined, as has the role of DNA damage in spore killing by moist heat and hydrogen peroxide. These data (1) suggest that core water content has little if any role in spore UV resistance and are consistent with binding of α/β-type SASP to DNA being the major mechanism providing protection to spores from UV radiation; (2) suggest that binding of αβ-type SASP to DNA is the major mechanism unique to spores providing protection from dry heat; (3) suggest that spore resistance to moist heat and hydrogen peroxide is affected to a large degree by the core water content, as increased core water resulted in large decreases in spore resistance to these agents; and (4) indicate that since this decreased resistance (i.e., in dacB spores) is not associated with increased spore killing by DNA damage, spore DNA must normally be extremely well protected against such damage, presumably by the saturation of spore DNA by α/β-type SASP. 19 refs., 2 figs., 5 tabs

  9. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

    1999-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  10. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  11. Crack Propagation Calculations for Optical Fibers under Static Bending and Tensile Loads Using Continuum Damage Mechanics

    Science.gov (United States)

    Chen, Yunxia; Cui, Yuxuan; Gong, Wenjun

    2017-01-01

    Static fatigue behavior is the main failure mode of optical fibers applied in sensors. In this paper, a computational framework based on continuum damage mechanics (CDM) is presented to calculate the crack propagation process and failure time of optical fibers subjected to static bending and tensile loads. For this purpose, the static fatigue crack propagation in the glass core of the optical fiber is studied. Combining a finite element method (FEM), we use the continuum damage mechanics for the glass core to calculate the crack propagation path and corresponding failure time. In addition, three factors including bending radius, tensile force and optical fiber diameter are investigated to find their impacts on the crack propagation process and failure time of the optical fiber under concerned situations. Finally, experiments are conducted and the results verify the correctness of the simulation calculation. It is believed that the proposed method could give a straightforward description of the crack propagation path in the inner glass core. Additionally, the predicted crack propagation time of the optical fiber with different factors can provide effective suggestions for improving the long-term usage of optical fibers. PMID:29140284

  12. Representation of dislocation cores using Nye tensor distributions

    International Nuclear Information System (INIS)

    Hartley, Craig S.; Mishin, Y.

    2005-01-01

    This paper demonstrates how the cores of atomistically simulated dislocations in Cu and Al can be represented by a distribution of infinitesimal dislocations described by appropriate components of the Nye tensor. Components calculated from atomic positions in the dislocated crystal are displayed as contour plots on the plane normal to the dislocation line. The method provides an accurate and instructive means for characterizing dislocation core structures and calculating the total Burgers vector

  13. The new definition of nuclear damage in the 1997 protocol to amend the 1963 vienna convention on civil liability for nuclear damage

    International Nuclear Information System (INIS)

    Soljan, V.

    2000-01-01

    This communication analyzes the content and the impact of the new definition of nuclear damage contented in the amendment protocol of the Vienna Convention relative to the civil liability in the 1963 Convention. Having in mind the experience of the Three Mile Island and Chernobyl accidents, it is demonstrated that the costs of preventive measures, damage to the environment and economic loss may constitute substantial portions of the total damage following a nuclear accident. Then, the new definition is studied in detail, on insisting on the notion of economic loss. A development is devoted to the question of damage to the environment. The preventive measures are studied and their conditions of the compensation receivability evoked with the criteria of reasonable measures. (N.C.)

  14. Instrumentation needs in LWR severe fuel damage experiments

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1980-01-01

    The Class 9 type nuclear accident is defined and the Three Mile Island type accident and proposed Idaho National Engineering Laboratory experiment series are described in some detail. Different types of severe fuel damage experiments are briefly discussed in order to show typical measurement requirements. General instrumentation needs and problems encountered in Class 9 accident research are outlined. It is concluded that the extremely high temperatures, high nuclear radiation fields, and oxidizing atmosphere will necessitate instrument development programs. Noncontact type sensing will be necessary in most of the molten core experiments

  15. Statistical damage constitutive model for rocks subjected to cyclic stress and cyclic temperature

    Science.gov (United States)

    Zhou, Shu-Wei; Xia, Cai-Chu; Zhao, Hai-Bin; Mei, Song-Hua; Zhou, Yu

    2017-10-01

    A constitutive model of rocks subjected to cyclic stress-temperature was proposed. Based on statistical damage theory, the damage constitutive model with Weibull distribution was extended. Influence of model parameters on the stress-strain curve for rock reloading after stress-temperature cycling was then discussed. The proposed model was initially validated by rock tests for cyclic stress-temperature and only cyclic stress. Finally, the total damage evolution induced by stress-temperature cycling and reloading after cycling was explored and discussed. The proposed constitutive model is reasonable and applicable, describing well the stress-strain relationship during stress-temperature cycles and providing a good fit to the test results. Elastic modulus in the reference state and the damage induced by cycling affect the shape of reloading stress-strain curve. Total damage induced by cycling and reloading after cycling exhibits three stages: initial slow increase, mid-term accelerated increase, and final slow increase.

  16. Review of fatigue criteria development for HTGR core supports

    International Nuclear Information System (INIS)

    Ho, F.H.; Vollman, R.E.

    1979-10-01

    Fatigue criteria for HTGR core support graphite structure are presented. The criteria takes into consideration the brittle nature of the material, and emphasizes the probabilistic approach in the treatment of strength data. The stress analysis is still deterministic. The conventional cumulative damage approach is adopted here. A specified minimum S-N curve is defined as the curve with 99% probability of survival at a 95% confidence level to accommodate random variability of the material strength. A constant life diagram is constructed to reconcile the effect of mean stress. The linear damage rule is assumed to account for the effect of random cycles. An additional factor of safety of three on cycles is recommended. The uniaxial S-N curve is modified in the medium-to-high cycle range (> 2 x 10 3 cycles) for mutiaxial fatigue effects

  17. PCBs and OCPs in sediment cores from the Lower St. Lawrence Estuary, Canada: evidence of fluvial inputs and time lag in delivery to coring sites.

    Science.gov (United States)

    Lebeuf, Michel; Nunes, Teresa

    2005-03-15

    Three sediment cores were collected along the longitudinal axis of the Laurentian Trough in the Lower St. Lawrence Estuary (LSLE) and an additional one at the junction of the Estuary and the Gulf of St. Lawrence. After core-slicing, each sediment layer was analyzed for polychlorinated biphenyls (PCBs) and some organochlorine pesticides (OCPs) including p,p'-dichlorodiphenyltrichloroethane (DDT) and its metabolites, hexachlorobenzene (HCB) and Mirex. 210Pb activity was also measured in these sediments, which allowed us to confirm that these cores were too much affected by the overall impact of surface mixing to be dated. Nevertheless, POP sedimentary profiles in cores from the LSLE upstream stations showed well-defined subsurface peak concentrations. Apparently, the peak inputs of POPs to these sediment cores had occurred after the years of maximum sales and production of these chemicals in North America, suggesting a time lag in the delivery of POPs to the LSLE sediments. Concentrations of POPs in the LSLE surface sediments as well as POP inventories in sediment cores decreased in the seaward direction, confirming that the head of the LSLE acts as a sink for sediments and associated constituents. Surface concentrations of sigmaPCBs, sigmaDDTs, and HCB in the most upstream core were on average similar to those reported in two fluvial lakes of the St. Lawrence River but were between 12 and 39 times lower than those from Lake Ontario. For Mirex, the surface concentration in that core was 5 and 130 times lower than the average values found in the fluvial lakes and Lake Ontario, respectively. Differences between Lake Ontario sediment cores and the most upstream core from the LSLE were much smaller on the basis of POP inventories than surface concentrations of POPs, but were still important. The total burdens of POPs in LSLE sediments below the 200 m isobath were 8704 kg for sigmaPCBs, 1825 kg for sigmaDDTs, 319 kg for HCB, and 27.5 kg for Mirex. These values represent

  18. Core-logs of the vertical borehole V2

    International Nuclear Information System (INIS)

    Carlsson, L.; Egerth, T.; Westlund, B.; Olsson, T.

    1982-08-01

    In the hydrogeological programme of the Stripa Project, borehole V2 was prolonged to a final depth of 822 m. The previous core from 0-471.4 m was relogged. The drill core was logged with regard to rock characteristics, fracture frequency, dipping and filling. The results are presented as core-logs and fracture diagrams. Borehole V2 shows similar characteristics as found in other drillings in the Stripa Mine. It penetrates Stripa granite to its full depth. recorded fractures shows a clear predominance of medium-steep fractures, while flat-lying fractures are more sparsly occuring, a fact which is even more pronounced below 400 m depth. Due to the vertical direction of the borehole, steeply dipping fractures are underestimated in the core. The mean fracture frequency, related to the total length of the core, is 2.1 fractures/m. Chlorite, calcite and epidote are the dominating coating minerals in the fractures, each making up about 25-30 percent of all coated fractures. (Authors)

  19. 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.

  20. Black bear damage to northwestern conifers in California: a review

    Science.gov (United States)

    Kenneth O. Fulgham; Dennis Hosack

    2017-01-01

    A total of 789 black bear damaged trees were investigate over a multi-year period on 14 different study sites chosen on lands of four participating timber companies. The sites ranged from 30 to 50 years of age. Four different conifer species were found to have black bear damage: coastal redwood (Sequoia sempervirens (D. Don) Endl.), Douglas-fir (...

  1. Core Stability and Core Selection in a Decentralized Labor Matching Market

    Directory of Open Access Journals (Sweden)

    Heinrich H. Nax

    2016-03-01

    Full Text Available We propose a dynamic model of decentralized many-to-one matching in the context of a competitive labor market. Through wage offers and wage demands, firms compete over workers and workers compete over jobs. Firms make hire-and-fire decisions dependent on the wages of their own workers and on the alternative workers available on the job market. Workers bargain for better jobs; either individually or collectively as unions, adjusting wage demands upward/downward depending on whether they are currently employed/unemployed. We show that such a process is absorbed into the core with probability one in finite time. Moreover, within the core, allocations are selected that are characterized by surplus splitting according to a bargaining solution such that (i firms and workforce share total revenue according to relative bargaining strengths, and (ii workers receive equal workforce shares above their individual outside options. These results bridge empirical evidence and provide a rich set of testable predictions.

  2. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  3. Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Malang, S.; Neitzel, H.J.

    1983-01-01

    Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

  4. Kombucha Tea Ameliorates Trichloroethylene Induced Hepatic Damages in Rats via Inhibition of Oxidative Stress and Free Radicals Induction

    International Nuclear Information System (INIS)

    Gharib, O.A.; Gharib, M.A.

    2008-01-01

    Kombucha Tea (KT) is reported to exhibit a wide variety of biological effects, including antioxidant. Evidence shows the important role of oxidative stress in the hepatic damage. The aim of this study is to investigate the possible protective effects of oral administration of KT in rats with trichloroethylene (TCE)-induced damage for ten consecutive days. Hepatic damage was evaluated by measuring total free radicals levels, biochemical and histological examinations. Serum gamma glutamyl transferase (GGT) activity (the hepatic damage marker), total protein, albumin and globulin as well as malonaldehyde (MDA), glutathione (GSH) content, nitric oxide (NO) concentration were evaluated in liver tissue homogenates. Total free radicals concentration in blood was examined by electron spin resonance (ESR). Total protein, DNA concentration, cell number and cell size in liver tissues were also examined. The rats orally administrated with TCE for ten days indicates hepatic damage changes, an increase in blood total free radicals concentration was observed, serum GGT activity, liver MDA, NO levels, total protein and decreased GSH content, DNA concentration and cell number. This accompanied with an increase in cell size of liver tissues, whereas KT reversed these effects. Furthermore, KT inhibits the concentration of total free radicals in blood and decreasing the increment of MDA and NO concentration. Histological studies reveal partial healing in those rats treated by KT after oral administration with TCE. The present results suggest that KT ameliorates TCE induced hepatic damage in rats probably due to its content of glucuronic, acetic acid and B vitamins via inhibition of oxidative stress and total free radicals

  5. Resumption of pulsing the NSCR following the discovery of damaged fuel

    International Nuclear Information System (INIS)

    Feltz, D.E.; Rogers, R.D.

    1984-01-01

    Pulsing operations of the Nuclear Science Center Reactor (NSCR) at Texas A and M University were terminated in 1976 following the discovery of three damaged fuel elements during a routine inspection. A commitment was then made to the U.S. Nuclear Regulatory Commission to terminate pulsing of the NSCR until a thorough study of the damaged fuel had been completed. A report describing that study and discussing the possible mechanism of damage was issued in 1981. Based on a recommendation in the report to establish a limiting temperature to protect against damage, the USNRC issued a letter authorizing the reinitiation of pulsing the NSCR but limiting pulsing parameters 'to those in the current technical specifications or to a maximum calculated fuel temperature of 830 deg. C. It is felt based on the data obtained and fuel inspection results that the requirements of Phase I and Phase III of the Pulse Test Program for Core VIII have been met. Phase II of the test program will not be implemented unless there is a requirement for higher pulse energy and flux. The reproducibility of pulse data was very satisfactory

  6. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  7. A systematic approach to the radiation damage problem in reactor materials

    International Nuclear Information System (INIS)

    Bullough, R.; Eyre, B.L.; Kulcinski, G.L.

    1976-09-01

    To assess the suitability of a material for use as a core component in a fast reactor or for the first wall in a fusion reactor, it is necessary to know the irradiation damage behaviour of the material outside the usual materials testing data domain. In the present paper a strategy is proposed based on a closely co-ordinated programme of experimental and theoretical research. The aim of this strategy is the systematic construction of a physically based model of the evolving damage structures. This would then allow both the necessary extrapolations of the data to the desired conditions to be achieved in a reliable fashion and provide a rational basis for the development of low swelling alloys for the two nuclear systems. (author)

  8. 3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

    Science.gov (United States)

    Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

    2017-10-01

    The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ funded by the US Department of Energy under Grant DE-SC0012315.

  9. Independent association of glucocorticoids with damage accrual in SLE.

    Science.gov (United States)

    Apostolopoulos, Diane; Kandane-Rathnayake, Rangi; Raghunath, Sudha; Hoi, Alberta; Nikpour, Mandana; Morand, Eric F

    2016-01-01

    To determine factors associated with damage accrual in a prospective cohort of patients with SLE. Patients with SLE who attended the Lupus Clinic at Monash Health, Australia, between 2007 and 2013 were studied. Clinical variables included disease activity (Systemic Lupus Erythematosus Disease Activity Index-2K, SLEDAI-2K), time-adjusted mean SLEDAI, cumulative glucocorticoid dose and organ damage (Systemic Lupus International Collaborating Clinics Damage Index (SDI)). Multivariate logistic regression analyses were performed to identify factors associated with damage accrual. A total of 162 patients were observed over a median (IQR) 3.6 (2.0-4.7) years. Seventy-five per cent (n=121) of patients received glucocorticoids. Damage accrual was significantly more frequent in glucocorticoid-exposed patients (42% vs 15%, p<0.01). Higher glucocorticoid exposure was independently associated with overall damage accrual after controlling for factors including ethnicity and disease activity and was significant at time-adjusted mean doses above 4.42 mg prednisolone/day; the OR of damage accrual in patients in the highest quartile of cumulative glucocorticoid exposure was over 10. Glucocorticoid exposure was independently associated with damage accrual in glucocorticoid-related and non-glucocorticoid related domains of the SDI. Glucocorticoid use is independently associated with the accrual of damage in SLE, including in non-glucocorticoid related domains.

  10. Structural organization of the quiescent core region in a turbulent channel flow

    International Nuclear Information System (INIS)

    Yang, Jongmin; Hwang, Jinyul; Sung, Hyung Jin

    2016-01-01

    Highlights: • The structural organization of the quiescent core region in a turbulent channel flow is explored. • The quiescent core region is the uniform momentum zone located at the center of the channel. • The boundary of the quiescent core region can be identified from the probability density function of the streamwise modal velocity. • The prograde and retrograde vortices form a counter-rotating vortex pair at the boundary of the core region. - Abstract: The structural organization of the quiescent core region in a turbulent channel flow was explored using direct numerical simulation data at Re_τ = 930. The quiescent core region is the uniform momentum zone located at the center of the channel, and contains the highest momentum with a low level of turbulence. The boundary of the quiescent core region can be identified from the probability density function of the streamwise modal velocity. The streamwise velocity changes abruptly near the boundary of the core region. The abrupt jump leads the increase of the velocity gradient, which is similar to the vorticity thickness of the laminar superlayer at the turbulent/non-turbulent interface. The strong shear induced from the abrupt change is originated from the vortical structure lying on the boundary of the core region. The spanwise population densities of the prograde and retrograde vortices have a local maximum near the boundary of the core region. The prograde vortex dominantly contributes to the total mean shear near the core boundary and the contribution to the total mean shear rapidly decreases within the core region. The prograde and retrograde vortices form a counter-rotating vortex pair at the boundary of the core region associated with the nibbling mechanism. The boundary of the core region contains large-scale concave and convex features. The concave (convex) core interface is organized by the negative-u (positive-u) regions which induce the ejections (sweeps) around the core boundary.

  11. Preliminary design of the new Proton Synchrotron Internal Dump core

    CERN Document Server

    AUTHOR|(CDS)2091975; Nuiry, François-Xavier

    The luminosity of the LHC particle accelerator at CERN is planned to be upgraded in the first half of 2020s, requiring also the upgrade of its injector accelerators, including the Proton Synchrotron (PS). The PS Internal Dumps are beam dumps located in the PS accelerator ring. They are safety devices designed to stop the circulating proton beam in order to protect the accelerator from damage due to an uncontrolled beam loss. The PS Internal Dumps need to be upgraded to be able to withstand the future higher intensity and energy proton beams. The dump core is a block of material interacting with the beam. It is located in ultra-high vacuum and moved into the beam path in 150 milliseconds by an electromagnet and spring-based actuation mechanism. The circulating proton beam is shaved by the core surface during thousands of beam revolutions. The preliminary new dump core design weighs 13 kilograms and consists of an isostatically pressed fine-grain graphite and a precipitation hardened copper alloy CuCrZr. The ...

  12. Development of Pipeline Database and CAD Model for Selection of Core Security Zone in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Choi, Seong Soo; Kwon, Tae Gyun; Baek, Hun Hyun; Kwon, Min Jin

    2008-07-01

    The objective of the project is to develop the pipeline database which can be used for selection of core security zones considering safety significance of pipes and to develop CAD model for 3-dimensional visualization of core security zones, for the purpose of minimizing damage and loss, enforcing security and protection on important facilities, and improving plant design preparing against emergency situations such as physical terrors in nuclear power plants. In this study, the pipeline database is developed for selection of core security zones considering safety significance of safety class 1 and 2 pipes. The database includes the information on 'pipe-room information-surrogate component' mapping, initiating events which may occur and accident mitigation functions which may be damaged by the pipe failure, and the drawing information related to 2,270 pipe segments of 30 systems. For the 3-dimensional visualization of core security zones, the CAD models on the containment building and the auxiliary building are developed using 3-D MAX tool and the demo program which can visualize the direct-X model converted from the 3-D MAX model is also developed. In addition to this, the coordinate information of all the buildings and their rooms is generated using AUTO CAD tool in order to be used as an input for 3-dimensional browsing of the VIP program

  13. On-line structural damage localization and quantification using wireless sensors

    International Nuclear Information System (INIS)

    Hsu, Ting-Yu; Huang, Shieh-Kung; Lu, Kung-Chung; Loh, Chin-Hsiung; Wang, Yang; Lynch, Jerome Peter

    2011-01-01

    In this paper, a wireless sensing system is designed to realize on-line damage localization and quantification of a structure using a frequency response function change method (FRFCM). Data interrogation algorithms are embedded in the computational core of the wireless sensing units to extract the necessary structural features, i.e. the frequency spectrum segments around eigenfrequencies, automatically from measured structural response for the FRFCM. Instead of the raw time history of the structural response, the extracted compact structural features are transmitted to the host computer. As a result, with less data transmitted from the wireless sensors, the energy consumed by the wireless transmission is reduced. To validate the performance of the proposed wireless sensing system, a six-story steel building with replaceable bracings in each story is instrumented with the wireless sensors for on-line damage detection during shaking table tests. The accuracy of the damage detection results using the wireless sensing system is verified through comparison with the results calculated from data recorded of a traditional wired monitoring system. The results demonstrate that, by taking advantage of collocated computing resources in wireless sensors, the proposed wireless sensing system can locate and quantify damage with acceptable accuracy and moderate energy efficiency

  14. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  15. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Budisantoso, Edi Trijono; Sardjono, Y.

    1996-01-01

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  16. Design criteria for a self-actuated shutdown system to ensure limitation of core damage

    International Nuclear Information System (INIS)

    Deane, N.A.; Atcheson, D.B.

    1981-09-01

    Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times

  17. What is the correlation of in vivo wear and damage patterns with in vitro TDR motion response?

    Science.gov (United States)

    Kurtz, Steven M.; Patwardhan, Avinash; MacDonald, Daniel; Ciccarelli, Lauren; van Ooij, André; Lorenz, Mark; Zindrick, Michael; O’Leary, Patrick; Isaza, Jorge; Ross, Raymond

    2008-01-01

    Background Context Total disc replacements (TDRs) have been used to reduce pain and preserve motion. However, the comparison of polyethylene wear following long-term implantation to those tested using an in vitro model had not yet been investigated. Purpose The purpose of this study was to correlate wear and damage patterns in retrieved TDRs with motion patterns observed in a clinically validated in vitro lumbar spine model. We also sought to determine whether one-sided wear and motion patterns were associated with greater in vivo wear. Study Design This two-part study combined the evaluation of retrieved total disc replacements with a biomechanical study using human lumbar spines. Patient Sample 38 CHARITÉ lumbar artificial discs were retrieved from 32 patients (24 female, 75%) after 7.3 years average implantation (range: 1.8 to 16.1y). The components were implanted at L2/L3 (n=1), L3/L4 (n=2), L4/L5 (n=20), and L5/S1 (n=15). All the implants were removed due to intractable back pain and/or facet degeneration. In addition, they were removed due to subsidence (n=10), anterior migration (n=3), core dislocation (n=2), lateral subluxation (n=1), endplate loosening (n = 2), and osteolysis (n=1). In parallel, 7 new implants were evaluated at L4-L5 and 13 implants at L5-S1 in an in vitro lumbar spine model. Outcome Measures Retrieval analysis included evaluation of clinical data, dimensional measurements and assessment of the extent and severity of PE surface damage mechanisms. In vitro testing involved the observation of motion patterns during physiological loading. Methods For the retrievals, each side of the PE core was independently analyzed at the rim and dome for the presence of machining marks, wear, and fracture. 35 cores were further analyzed using MicroCT to determine whether the wear was one-sided, or symmetrically distributed. For the in vitro study the new implants were tested under physiologic loads (flexion-extension with a compressive follower preload

  18. Damage on sliding bearings of internal combustion engines. Damage patterns, causes, prevention; Schaeden an Gleitlagern von Verbrennungsmotoren. Erscheinungsbilder, Ursachen, Vermeidung

    Energy Technology Data Exchange (ETDEWEB)

    Ederer, U.G. [Miba Gleitlager GmbH, Laakrichen (Austria)

    2005-07-01

    Bearing failures are consequences of system deficiencies which cause an inadequate function of the hydrodynamic action and, thereby, too high a friction, at least locally. The bearing overheats, what ultimately leads to its destruction and that of adjacent components. These 'consequential damages' are frequently severe. We identify, therefore, early stages of malfunction, already as 'bearing damage'. In this condition, a diagnosis and remedial measures to avoid total destruction are possible. Typical bearing conditions, possible causes and remedies are described herein. (orig.)

  19. AKT phosphorylates H3-threonine 45 to facilitate termination of gene transcription in response to DNA damage.

    Science.gov (United States)

    Lee, Jong-Hyuk; Kang, Byung-Hee; Jang, Hyonchol; Kim, Tae Wan; Choi, Jinmi; Kwak, Sojung; Han, Jungwon; Cho, Eun-Jung; Youn, Hong-Duk

    2015-05-19

    Post-translational modifications of core histones affect various cellular processes, primarily through transcription. However, their relationship with the termination of transcription has remained largely unknown. In this study, we show that DNA damage-activated AKT phosphorylates threonine 45 of core histone H3 (H3-T45). By genome-wide chromatin immunoprecipitation sequencing (ChIP-seq) analysis, H3-T45 phosphorylation was distributed throughout DNA damage-responsive gene loci, particularly immediately after the transcription termination site. H3-T45 phosphorylation pattern showed close-resemblance to that of RNA polymerase II C-terminal domain (CTD) serine 2 phosphorylation, which establishes the transcription termination signal. AKT1 was more effective than AKT2 in phosphorylating H3-T45. Blocking H3-T45 phosphorylation by inhibiting AKT or through amino acid substitution limited RNA decay downstream of mRNA cleavage sites and decreased RNA polymerase II release from chromatin. Our findings suggest that AKT-mediated phosphorylation of H3-T45 regulates the processing of the 3' end of DNA damage-activated genes to facilitate transcriptional termination. © The Author(s) 2015. Published by Oxford University Press on behalf of Nucleic Acids Research.

  20. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

  1. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis

  2. Building Damage Extraction Triggered by Earthquake Using the Uav Imagery

    Science.gov (United States)

    Li, S.; Tang, H.

    2018-04-01

    When extracting building damage information, we can only determine whether the building is collapsed using the post-earthquake satellite images. Even the satellite images have the sub-meter resolution, the identification of slightly damaged buildings is still a challenge. As the complementary data to satellite images, the UAV images have unique advantages, such as stronger flexibility and higher resolution. In this paper, according to the spectral feature of UAV images and the morphological feature of the reconstructed point clouds, the building damage was classified into four levels: basically intact buildings, slightly damaged buildings, partially collapsed buildings and totally collapsed buildings, and give the rules of damage grades. In particular, the slightly damaged buildings are determined using the detected roof-holes. In order to verify the approach, we conduct experimental simulations in the cases of Wenchuan and Ya'an earthquakes. By analyzing the post-earthquake UAV images of the two earthquakes, the building damage was classified into four levels, and the quantitative statistics of the damaged buildings is given in the experiments.

  3. Interstitial pO2 in ischemic penumbra and core are differentially affected following transient focal cerebral ischemia in rats.

    Science.gov (United States)

    Liu, Shimin; Shi, Honglian; Liu, Wenlan; Furuichi, Takamitsu; Timmins, Graham S; Liu, Ke Jian

    2004-03-01

    Stroke causes heterogeneous changes in tissue oxygenation, with a region of decreased blood flow, the penumbra, surrounding a severely damaged ischemic core. Treatment of acute ischemic stroke aims to save this penumbra before its irreversible damage by continued ischemia. However, effective treatment remains elusive due to incomplete understanding of processes leading to penumbral death. While oxygenation is central in ischemic neuronal death, it is unclear exactly what actual changes occur in interstitial oxygen tension (pO2) in ischemic regions during stroke, particularly the penumbra. Using the unique capability of in vivo electron paramagnetic resonance (EPR) oximetry to measure localized interstitial pO2, we measured both absolute values, and temporal changes of pO2 in ischemic penumbra and core during ischemia and reperfusion in a rat model. Ischemia rapidly decreased interstitial pO2 to 32% +/- 7.6% and 4% +/- 0.6% of pre-ischemic values in penumbra and core, respectively 1 hour after ischemia. Importantly, whilst reperfusion restored core pO2 close to its pre-ischemic value, penumbral pO2 only partially recovered. Hyperoxic treatment significantly increased penumbral pO2 during ischemia, but not in the core, and also increased penumbral pO2 during reperfusion. These divergent, important changes in pO2 in penumbra and core were explained by combined differences in cellular oxygen consumption rates and microcirculation conditions. We therefore demonstrate that interstitial pO2 in penumbra and core is differentially affected during ischemia and reperfusion, providing new insights to the pathophysiology of stroke. The results support normobaric hyperoxia as a potential early intervention to save penumbral tissue in acute ischemic stroke.

  4. Improvement of open and semi-open core wall system in tall buildings by closing of the core section in the last story

    Science.gov (United States)

    Kheyroddin, A.; Abdollahzadeh, D.; Mastali, M.

    2014-09-01

    Increasing number of tall buildings in urban population caused development of tall building structures. One of the main lateral load resistant systems is core wall system in high-rise buildings. Core wall system has two important behavioral aspects where the first aspect is related to reduce the lateral displacement by the core bending resistance and the second is governed by increasing of the torsional resistance and core warping of buildings. In this study, the effects of closed section core in the last story have been considered on the behavior of models. Regarding this, all analyses were performed by ETABS 9.2.v software (Wilson and Habibullah). Considering (a) drift and rotation of the core over height of buildings, (b) total and warping stress in the core body, (c) shear in beams due to warping stress, (d) effect of closing last story on period of models in various modes, (e) relative displacement between walls in the core system and (f) site effects in far and near field of fault by UBC97 spectra on base shear coefficient showed that the bimoment in open core is negative in the last quarter of building and it is similar to wall-frame structures. Furthermore, analytical results revealed that closed section core in the last story improves behavior of the last quarter of structure height, since closing of core section in the last story does not have significant effect on reducing base shear value in near and far field of active faults.

  5. NDR1 modulates the UV-induced DNA-damage checkpoint and nucleotide excision repair

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong-Min; Choi, Ji Ye [Department of Biological Science, Dong-A University, Busan (Korea, Republic of); Yi, Joo Mi [Research Center, Dongnam Institute of Radiological & Medical Sciences, Busan (Korea, Republic of); Chung, Jin Woong; Leem, Sun-Hee; Koh, Sang Seok [Department of Biological Science, Dong-A University, Busan (Korea, Republic of); Kang, Tae-Hong, E-mail: thkang@dau.ac.kr [Department of Biological Science, Dong-A University, Busan (Korea, Republic of)

    2015-06-05

    Nucleotide excision repair (NER) is the sole mechanism of UV-induced DNA lesion repair in mammals. A single round of NER requires multiple components including seven core NER factors, xeroderma pigmentosum A–G (XPA–XPG), and many auxiliary effector proteins including ATR serine/threonine kinase. The XPA protein helps to verify DNA damage and thus plays a rate-limiting role in NER. Hence, the regulation of XPA is important for the entire NER kinetic. We found that NDR1, a novel XPA-interacting protein, modulates NER by modulating the UV-induced DNA-damage checkpoint. In quiescent cells, NDR1 localized mainly in the cytoplasm. After UV irradiation, NDR1 accumulated in the nucleus. The siRNA knockdown of NDR1 delayed the repair of UV-induced cyclobutane pyrimidine dimers in both normal cells and cancer cells. It did not, however, alter the expression levels or the chromatin association levels of the core NER factors following UV irradiation. Instead, the NDR1-depleted cells displayed reduced activity of ATR for some set of its substrates including CHK1 and p53, suggesting that NDR1 modulates NER indirectly via the ATR pathway. - Highlights: • NDR1 is a novel XPA-interacting protein. • NDR1 accumulates in the nucleus in response to UV irradiation. • NDR1 modulates NER (nucleotide excision repair) by modulating the UV-induced DNA-damage checkpoint response.

  6. Develop a practical means to monitor the criticality of the TMI-2 core

    International Nuclear Information System (INIS)

    Kim, S.S.; Levine, S.H.; Imel, G.

    1984-06-01

    A method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. It is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs the nucleonic codes to analyze in two dimensions the neutron multiplication of a Cf-252 neutron source. Experiments were performed on the Penn State Breazeale TRIGA Reactor (PSBR). The first set of experiments calibrates the k/sub infinity/'s of the fuel elements moved during the second set of experiments. The second set of experiments provides a means for both developing and validating the AMPNS method. Several test runs of optimization calculations have been made on the PSBR core assuming one of the subcritical configurations is a damaged core. Test runs of the AMPNS method reveals that when the core cell size and source position are correctly chosen, the solution converges to the correct k/sub eff/ and k/sub infinity/ distribution without any oscillations or instabilities. Application of the AMPNS method to the degraded TMI-2 core has been studied to provide some initial insight into this 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. Core-to-core uniformity improvement in multi-core fiber Bragg gratings

    Science.gov (United States)

    Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

    2014-07-01

    Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

  9. Probabilistic Assessment of Structural Seismic Damage for Buildings in Mid-America

    International Nuclear Information System (INIS)

    Bai, Jong-Wha; Hueste, Mary Beth D.; Gardoni, Paolo

    2008-01-01

    This paper provides an approach to conduct a probabilistic assessment of structural damage due to seismic events with an application to typical building structures in Mid-America. The developed methodology includes modified damage state classifications based on the ATC-13 and ATC-38 damage states and the ATC-38 database of building damage. Damage factors are assigned to each damage state to quantify structural damage as a percentage of structural replacement cost. To account for the inherent uncertainties, these factors are expressed as random variables with a Beta distribution. A set of fragility curves, quantifying the structural vulnerability of a building, is mapped onto the developed methodology to determine the expected structural damage. The total structural damage factor for a given seismic intensity is then calculated using a probabilistic approach. Prediction and confidence bands are also constructed to account for the prevailing uncertainties. The expected seismic structural damage is assessed for a typical building structure in the Mid-America region using the developed methodology. The developed methodology provides a transparent procedure, where the structural damage factors can be updated as additional seismic damage data becomes available

  10. LMFR core thermohydraulics: Status and prospects

    International Nuclear Information System (INIS)

    2000-06-01

    One of the fundamental steps for a successful reactor core thermohydraulic design is the capability to predict, reliably and accurately, the temperature distribution in the core assemblies. A detailed knowledge of the assembly and fuel pin thermohydraulic behaviour in the steady state and transient conditions is an indispensable prerequisite to safe and stable operation of the reactor. Considerable experimental and theoretical studies on various aspects of LMFR core thermohydraulics are necessary to acquire such knowledge. During the last decade, there have been substantial advances in fast reactor core thermohydraulic design and operation in several countries with fast reactor programmes (notably in France, the Russian Federation, Japan, the United Kingdom, Germany and the United States of America). Chief among these has been the demonstration of reliable operation of reactor cores at a high burnup. During the last years, some additional countries such as China, India and the Republic of Korea have launched new fast reactor programmes. International exchange of information and experience on LMFR development including core thermohydraulic design is becoming of increasing importance to these countries. It is with this focus that the IAEA convened the Technical Committee on 'Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR's with Traditional and Burner Cores'. This meeting, attended by participants from seven countries, brought together a group of international experts to review and discuss the thermohydraulic advances and design approaches providing a reliable, safe and robust reactor core, as well as to exchange the experience accumulated in different countries of using the codes for thermohydraulic calculations and to discuss the issues requiring further research and development. A total of thirty technical papers presented covered theoretical and computational issues as well as experiments under

  11. Hardware concepts for a large low-energetics LMFBR core. Final report

    International Nuclear Information System (INIS)

    Hutter, E.; Batch, R.V.

    1980-12-01

    A design study was made to identify a practical set of hardware configurations that would embody the requirements developed in the numerical study of a low-energetics core and blanket for a prototype large breeder reactor. Dimensioned drawings are presented for fuel, blanket, reflector/shield, and control rod subassemblies. A horizontal cross section drawing shows how these subassemblies are arranged in the total core/blanket assembly. A core support is illustrated showing a dual plenums arrangement

  12. Deep rock damage in the San Andreas Fault revealed by P- and S-type fault-zone-guided waves

    Science.gov (United States)

    Ellsworth, William L.; Malin, Peter E.

    2011-01-01

    Damage to fault-zone rocks during fault slip results in the formation of a channel of low seismic-wave velocities. Within such channels guided seismic waves, denoted by Fg, can propagate. Here we show with core samples, well logs and Fg-waves that such a channel is crossed by the SAFOD (San Andreas Fault Observatory at Depth) borehole at a depth of 2.7 km near Parkfield, California, USA. This laterally extensive channel extends downwards to at least half way through the seismogenic crust, more than about 7 km. The channel supports not only the previously recognized Love-type- (FL) and Rayleigh-type- (FR) guided waves, but also a new fault-guided wave, which we name FF. As recorded 2.7 km underground, FF is normally dispersed, ends in an Airy phase, and arrives between the P- and S-waves. Modelling shows that FF travels as a leaky mode within the core of the fault zone. Combined with the drill core samples, well logs and the two other types of guided waves, FF at SAFOD reveals a zone of profound, deep, rock damage. Originating from damage accumulated over the recent history of fault movement, we suggest it is maintained either by fracturing near the slip surface of earthquakes, such as the 1857 Fort Tejon M 7.9, or is an unexplained part of the fault-creep process known to be active at this site.

  13. An extended sequence specificity for UV-induced DNA damage.

    Science.gov (United States)

    Chung, Long H; Murray, Vincent

    2018-01-01

    The sequence specificity of UV-induced DNA damage was determined with a higher precision and accuracy than previously reported. UV light induces two major damage adducts: cyclobutane pyrimidine dimers (CPDs) and pyrimidine(6-4)pyrimidone photoproducts (6-4PPs). Employing capillary electrophoresis with laser-induced fluorescence and taking advantages of the distinct properties of the CPDs and 6-4PPs, we studied the sequence specificity of UV-induced DNA damage in a purified DNA sequence using two approaches: end-labelling and a polymerase stop/linear amplification assay. A mitochondrial DNA sequence that contained a random nucleotide composition was employed as the target DNA sequence. With previous methodology, the UV sequence specificity was determined at a dinucleotide or trinucleotide level; however, in this paper, we have extended the UV sequence specificity to a hexanucleotide level. With the end-labelling technique (for 6-4PPs), the consensus sequence was found to be 5'-GCTC*AC (where C* is the breakage site); while with the linear amplification procedure, it was 5'-TCTT*AC. With end-labelling, the dinucleotide frequency of occurrence was highest for 5'-TC*, 5'-TT* and 5'-CC*; whereas it was 5'-TT* for linear amplification. The influence of neighbouring nucleotides on the degree of UV-induced DNA damage was also examined. The core sequences consisted of pyrimidine nucleotides 5'-CTC* and 5'-CTT* while an A at position "1" and C at position "2" enhanced UV-induced DNA damage. Crown Copyright © 2017. Published by Elsevier B.V. All rights reserved.

  14. In situ observation of modulated light emission of fiber fuse synchronized with void train over hetero-core splice point.

    Directory of Open Access Journals (Sweden)

    Shin-ichi Todoroki

    Full Text Available BACKGROUND: Fiber fuse is a process of optical fiber destruction under the action of laser radiation, found 20 years ago. Once initiated, opical discharge runs along the fiber core region to the light source and leaves periodic voids whose shape looks like a bullet pointing the direction of laser beam. The relation between damage pattern and propagation mode of optical discharge is still unclear even after the first in situ observation three years ago. METHODOLOGY/PRINCIPAL FINDINGS: Fiber fuse propagation over hetero-core splice point (Corning SMF-28e and HI 1060 was observed in situ. Sequential photographs obtained at intervals of 2.78 micros recorded a periodic emission at the tail of an optical discharge pumped by 1070 nm and 9 W light. The signal stopped when the discharge ran over the splice point. The corresponding damage pattern left in the fiber core region included a segment free of periodicity. CONCLUSIONS: The spatial modulation pattern of the light emission agreed with the void train formed over the hetero-core splice point. Some segments included a bullet-shaped void pointing in the opposite direction to the laser beam propagation although the sequential photographs did not reveal any directional change in the optical discharge propagation.

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

    International Nuclear Information System (INIS)

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

    1997-03-01

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

  16. Analysis of Boling's laser-damage morphology

    International Nuclear Information System (INIS)

    Sparks, M.S.

    1980-01-01

    Boling observed that his total-internal-reflection laser-damage sites in glass closely resembled the scattering cross section for small (ka << 1), perfectly conducting sphere and suggested that a very small plasma formed and grew to a larger size, still with ka << 1 satisfied. Even with ka = 1, for which the cross section is different from that observed, the scattered field still is too small to explain the damage in terms of constructive interference between the incident- and scattered fields. Furthermore, the characteristic shape of the scattering cross section that matches the damage patterns is for circular polarization or unpolarized light, in contrast to the experimental plane polarizations. Extending the ideas to include effects of the scattered field outside the glass, such as plasma formation, and to include the correct field (with interesting polarization, including longitudinal circuler polarization at certain distances from the surface) incident on the sphere may explain the experiments. Additional experiments and analysis would be useful to determine if the extended model is valid and to investigate related materials improvement, nondestructive testing, and the relation between laser damage, plasma initiation, and failure under stress, all initiated at small isolated spots

  17. Diffusional mass transport phenomena in the buffer material and damaged zone of a borehole wall in an underground nuclear fuel waste vault

    International Nuclear Information System (INIS)

    Page, S.; Cheung, S.C.H.

    1983-06-01

    The effects of the geometry of the borehole and the characteristics of the damaged borehole rock wall on the movement of the radionuclides from an underground nuclear waste vault have been studied. The results show that radionuclide transport will occur mainly through the buffer into the damaged zone of the borehole wall. As the degree of facturing of the damaged zone increases, the total radionuclide flux will increase up to a limit which can be approximated by a one-dimensional radial diffusion model. For large degrees of fracturing of the damaged zone, an increase in the radial buffer material thickness will decrease the total flux, whereas, for small degrees of fracturing, an increase in the radial buffer thickness may slightly increase the total flux. Increasing the vertical buffer thickness will significantly decrease the total flux when the degree of fracturing of the damaged zone is small. An increase in the vertical extent of the damaged zone will cause an increase in total flux

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

    Science.gov (United States)

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

    2008-11-01

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

  19. Report of the working group for nuclear damage compensation system

    International Nuclear Information System (INIS)

    1989-01-01

    The Working Group for Nuclear Damage Compensation System was established within the Atomic Energy Commision of Japan on August 2, 1988. The Group has held five meetings to make a study on the revision of the reserve for nuclear damage compensation. The nuclear damage compensation system in Japan has been established under the Law Concerning Compensation for Nuclear Damages and the Law Concerning Contract for Compensation for Nuclear Damages. The former law requires the nuclear power plant operators to set up a reserve for damage compensation to ensure positive and quick payment of compensation in the event of an accident. The reserve is currently rely on liability insurance and a government compensation contract. The Working Group has concluded that the total reserve should be increased from the current yen10 bill. to yen30 bill. The amount of the reserve specified in the enforcement law for the Law Concerning Compensation for Nuclear Damages should also be increased accordingly. The Law Concerning compensation for Nuclear damage will also be applied to damage which occurs overseas as a result of an accident in Japan. (N.K.)

  20. Dose rate effects during damage accumulation in silicon

    Energy Technology Data Exchange (ETDEWEB)

    Caturla, M.J.; Diaz de la Rubia, T.

    1997-01-01

    We combine molecular dynamics and Monte Carlo simulations to study damage accumulation and dose rate effects during irradiation of Silicon. We obtain the initial stage of the damage produced by heavy and light ions using classical molecular dynamics simulations. While heavy ions like As or Pt induce amorphization by single ion impact, light ions like B only produce point defects or small clusters of defects. The amorphous pockets generated by heavy ions are stable below room temperature and recrystallize at temperatures below the threshold for recrystallization of a planar amorphous-crystalline interface. The damage accumulation during light ion irradiation is simulated using a Monte Carlo model for defect diffusion. In this approach, we study the damage in the lattice as a function of dose and dose rate. A strong reduction in the total number of defects left in the lattice is observed for lower dose rates.

  1. Dose rate effects during damage accumulation in silicon

    International Nuclear Information System (INIS)

    Caturla, M.J.; Diaz de la Rubia, T.

    1997-01-01

    The authors combine molecular dynamics and Monte Carlo simulations to study damage accumulation and dose rate effects during irradiation of silicon. They obtain the initial stage of the damage produced by heavy and light ions using classical molecular dynamics simulations. While heavy ions like As or Pt induce amorphization by single ion impact, light ions like B only produce point defects or small clusters of defects. The amorphous pockets generated by heavy ions are stable below room temperature and recrystallize at temperatures below the threshold for recrystallization of a planar amorphous-crystalline interface. The damage accumulation during light ion irradiation is simulated using a Monte Carlo model for defect diffusion. In this approach, the authors study the damage in the lattice as a function of dose and dose rate. A strong reduction in the total number of defects left in the lattice is observed for lower dose rates

  2. Coupling between core and cladding modes in a helical core fiber with large core offset

    International Nuclear Information System (INIS)

    Napiorkowski, Maciej; Urbanczyk, Waclaw

    2016-01-01

    We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

  3. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    International Nuclear Information System (INIS)

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and 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 safely to a full-silicide core

  4. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  5. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

    Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

  6. Metric to quantify white matter damage on brain magnetic resonance images

    International Nuclear Information System (INIS)

    Valdes Hernandez, Maria del C.; Munoz Maniega, Susana; Anblagan, Devasuda; Bastin, Mark E.; Wardlaw, Joanna M.; Chappell, Francesca M.; Morris, Zoe; Sakka, Eleni; Dickie, David Alexander; Royle, Natalie A.; Armitage, Paul A.; Deary, Ian J.

    2017-01-01

    Quantitative assessment of white matter hyperintensities (WMH) on structural Magnetic Resonance Imaging (MRI) is challenging. It is important to harmonise results from different software tools considering not only the volume but also the signal intensity. Here we propose and evaluate a metric of white matter (WM) damage that addresses this need. We obtained WMH and normal-appearing white matter (NAWM) volumes from brain structural MRI from community dwelling older individuals and stroke patients enrolled in three different studies, using two automatic methods followed by manual editing by two to four observers blind to each other. We calculated the average intensity values on brain structural fluid-attenuation inversion recovery (FLAIR) MRI for the NAWM and WMH. The white matter damage metric is calculated as the proportion of WMH in brain tissue weighted by the relative image contrast of the WMH-to-NAWM. The new metric was evaluated using tissue microstructure parameters and visual ratings of small vessel disease burden and WMH: Fazekas score for WMH burden and Prins scale for WMH change. The correlation between the WM damage metric and the visual rating scores (Spearman ρ > =0.74, p =0.72, p < 0.0001). The repeatability of the WM damage metric was better than WM volume (average median difference between measurements 3.26% (IQR 2.76%) and 5.88% (IQR 5.32%) respectively). The follow-up WM damage was highly related to total Prins score even when adjusted for baseline WM damage (ANCOVA, p < 0.0001), which was not always the case for WMH volume, as total Prins was highly associated with the change in the intense WMH volume (p = 0.0079, increase of 4.42 ml per unit change in total Prins, 95%CI [1.17 7.67]), but not with the change in less-intense, subtle WMH, which determined the volumetric change. The new metric is practical and simple to calculate. It is robust to variations in image processing methods and scanning protocols, and sensitive to subtle and severe white

  7. Permeability of sediment cores from methane hydrate deposit in the Eastern Nankai Trough, Japan

    Science.gov (United States)

    Konno, Y.; Yoneda, J.; Egawa, K.; Ito, T.; Jin, Y.; Kida, M.; Suzuki, K.; Nakatsuka, Y.; Nagao, J.

    2013-12-01

    Effective and absolute permeability are key parameters for gas production from methane-hydrate-bearing sandy sediments. Effective and/or absolute permeability have been measured using methane-hydrate-bearing sandy cores and clayey and silty cores recovered from Daini Atsumi Knoll in the Eastern Nankai Trough during the 2012 JOGMEC/JAPEX Pressure coring operation. Liquid-nitrogen-immersed cores were prepared by rapid depressurization of pressure cores recovered by a pressure coring system referred to as the Hybrid PCS. Cores were shaped cylindrically on a lathe with spraying of liquid nitrogen to prevent hydrate dissociation. Permeability was measured by a flooding test or a pressure relaxation method under near in-situ pressure and temperature conditions. Measured effective permeability of hydrate-bearing sediments is less than tens of md, which are order of magnitude less than absolute permeability. Absolute permeability of clayey cores is approximately tens of μd, which would perform a sealing function as cap rocks. Permeability reduction due to a swelling effect was observed for a silty core during flooding test of pure water mimicking hydrate-dissociation-water. Swelling effect may cause production formation damage especially at a later stage of gas production from methane hydrate deposits. This study was financially supported by the Research Consortium for Methane Hydrate Resources in Japan (MH21 Research Consortium) that carries out Japan's Methane Hydrate R&D Program conducted by the Ministry of Economy, Trade and Industry (METI).

  8. A non-algorithmic approach to the In-core-fuel management problem of a PWR core

    International Nuclear Information System (INIS)

    Kimhy, Y.

    1992-03-01

    The primary objective of a commercial nuclear power plant operation is to produce electricity a low cost while satisfying safety constraints imposed on the operating conditions. Design of a fuel reload cycle for the current generation nuclear power plant represents a multistage process with a series of design decisions taken at various time points. Of these stages, reload core design is an important stage, due to its impact on safety and economic plant performance parameters. Overall. performance of the plant during the power production cycle depends on chosen fresh fuel parameters, as well as specific fuel configuration of the reactor core. The motivation to computerize generation and optimization of fuel reload configurations follows from some reasons: first, reload is performed periodically and requires manipulation of a large amount of data. second, in recent years, more complicated fuel loading patterns were developed and implemented following changes in fuel design and/or operational requirements, such as, longer cycles, advanced burnable poison designs, low leakage loading patterns and reduction of irradiation-induced damage of the pressure vessel. An algorithmic approach to the problem was generally adopted. The nature of the reload design process is a 'heuristic' search performed manually by a fuel manager. The knowledge used by the fuel manager is mostly accumulated experience in reactor physics and core calculations. These features of the problem and the inherent disadvantage of the algorithmic method are the main reasons to explore a non-algorithmic approach for solving the reload configuration problem. Several features of the 'solutions space' ( a collection of acceptable final configurations ) are emphasized in this work: 1) the space contain numerous number of entities (> 25) that are distributed un homogeneously, 2) the lack of a monotonic objective function decrease the probability to find an isolated optimum configuration by depth first search or

  9. Selective logging and damage to unharvested trees in a hyrcanian forest of Iran

    OpenAIRE

    Farshad Keivan Behjou; Omid Ghafarzade Mollabashi

    2012-01-01

    Selective logging in mature hardwood stands of Caspian forests often causes physical damage to residual trees through felling and skidding operations, resulting in a decline in bole quality and subsequent loss of tree value. This study evaluated the logging damage to residual trees following logging operations. A total density of 5.1 trees/ha and 17.3 m3/ha of wood were harvested. On average, 9.8 trees were damaged for every tree extracted, including 8 trees destroyed or severely damaged. The...

  10. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  11. Damage analysis: damage function development and application

    International Nuclear Information System (INIS)

    Simons, R.L.; Odette, G.R.

    1975-01-01

    The derivation and application of damage functions, including recent developments for the U.S. LMFBR and CTR programs, is reviewed. A primary application of damage functions is in predicting component life expectancies; i.e., the fluence required in a service spectrum to attain a specified design property change. An important part of the analysis is the estimation of the uncertainty in such fluence limit predictions. The status of standardizing the procedures for the derivation and application of damage functions is discussed. Improvements in several areas of damage function development are needed before standardization can be completed. These include increasing the quantity and quality of the data used in the analysis, determining the limitations of the analysis due to the presence of multiple damage mechanisms, and finally, testing of damage function predictions against data obtained from material surveillance programs in operating thermal and fast reactors. 23 references. (auth)

  12. Analysis of core damage frequency from internal events: Expert judgment elicitation. Part 1: Expert panel results. Part 2: Project staff results

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, T A; Cramond, W R [Sandia National Laboratories, Albuquerque, NM (United States); Hora, S C [University of Hawii at Hilo (United States); Unwin, S D [Brookhaven National Laboratory (United States)

    1989-04-01

    Quantitative modeling techniques have limitations as to the resolution of important issues in probabilistic risk assessment (PRA). Not all issues can be resolved via the existing set of methods such as fault trees, event trees, statistical analyses, data collection, and computer simulation. Therefore, an expert judgment process was developed to address issues perceived as important to risk in the NUREG-1150 analysis but which could not be resolved with existing techniques. This process was applied to several issues that could significantly affect the internal event core damage frequencies of the PRAs performed on six light water reactors. Detailed descriptions of these issues and the results of the expert judgment elicitation are reported here, as well as an explanation of the methodology used and the procedure followed in performing the overall elicitation task. The process is time-consuming and expensive. However, the results are very useful, and represent an improvement over the draft NUREG-1150 analysis in the areas of expert selection, elicitation training, issue selection and presentation, elicitation of judgment and aggregation of results. The results are presented in two parts. Part documents the expert panel elicitations, where the most important issues were presented to a panel of experts convened from throughout the nuclear power risk assessment community. Part 2 documents the process by which the project staff performed expert judgment on other important issues, using the project staff as panel members. (author)

  13. Spectrum measurements in the ZENITH plutonium core 7 using a neutron chopper

    Energy Technology Data Exchange (ETDEWEB)

    Barclay, F R; Cameron, I R; Pitcher, H H.W.; Symons, C R [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-05-15

    As part of the experimental programme on the first plutonium loading of ZENITH (Core 7) a series of measurements was carried out with the neutron chopper on a beam emerging from the core centre. The general experimental programme on the two ZENITH plutonium cores has been covered elsewhere. Core 7 had a carbon/Pu239 atomic ratio of 2666 and a steel/Pu239 ratio of 76.8, giving an absorption cross-section at 2200 m/sec. of 0.31 barns/carbon atom. The fuel was in the form of 'spikes' of 0.020 in. thick Pu/Al alloy sheathed in 0.020 in. aluminium, the isotopic composition of the plutonium being 97.4% Pu239, 2.55% Pu240 and 0.1% Pu241. The overall layout of the reactor core and reflector is shown in the vertical section through the reactor vessel and the plan view. The core consists of a vertical array of 235 cylindrical graphite sleeves of outer diameter 7.37 cm into each of which a cylindrical graphite box may be loaded. Sunning longitudinally inside the box are six parallel grooves which act as locations for the edges of either the Pu/Al spikes or graphite dummies of the same external dimensions. Each groove accommodates two spikes end-to-end, with a small graphite spacer between to avoid welding together of the spike sheaths when heated. Lateral spacers of graphite or stainless steel fill the five spaces between the six spikes or dummies. The total length of the plutonium-loaded core region is 140 cm, the ends of the element forming graphite reflectors of length 53 cm. In Core 7 each fuel element contained 10 Pu-Al spikes. The fuel elements are arranged in a triangular lattice of pitch 7.62 cm to form the reactor core, of diameter 1.23 m. A radial graphite reflector approximately 1 metre thick surrounds the core and is separated from it by an annular lampblack thermal barrier, contained within graphite tiles, which reduces heat transfer from the core. The reactor can be heated by circulation of nitrogen through a 250 kW heater below the core. The nitrogen flows

  14. Spectrum measurements in the ZENITH plutonium core 7 using a neutron chopper

    International Nuclear Information System (INIS)

    Barclay, F.R.; Cameron, I.R.; Pitcher, H.H.W.; Symons, C.R.

    1964-05-01

    As part of the experimental programme on the first plutonium loading of ZENITH (Core 7) a series of measurements was carried out with the neutron chopper on a beam emerging from the core centre. The general experimental programme on the two ZENITH plutonium cores has been covered elsewhere. Core 7 had a carbon/Pu239 atomic ratio of 2666 and a steel/Pu239 ratio of 76.8, giving an absorption cross-section at 2200 m/sec. of 0.31 barns/carbon atom. The fuel was in the form of 'spikes' of 0.020 in. thick Pu/Al alloy sheathed in 0.020 in. aluminium, the isotopic composition of the plutonium being 97.4% Pu239, 2.55% Pu240 and 0.1% Pu241. The overall layout of the reactor core and reflector is shown in the vertical section through the reactor vessel and the plan view. The core consists of a vertical array of 235 cylindrical graphite sleeves of outer diameter 7.37 cm into each of which a cylindrical graphite box may be loaded. Sunning longitudinally inside the box are six parallel grooves which act as locations for the edges of either the Pu/Al spikes or graphite dummies of the same external dimensions. Each groove accommodates two spikes end-to-end, with a small graphite spacer between to avoid welding together of the spike sheaths when heated. Lateral spacers of graphite or stainless steel fill the five spaces between the six spikes or dummies. The total length of the plutonium-loaded core region is 140 cm, the ends of the element forming graphite reflectors of length 53 cm. In Core 7 each fuel element contained 10 Pu-Al spikes. The fuel elements are arranged in a triangular lattice of pitch 7.62 cm to form the reactor core, of diameter 1.23 m. A radial graphite reflector approximately 1 metre thick surrounds the core and is separated from it by an annular lampblack thermal barrier, contained within graphite tiles, which reduces heat transfer from the core. The reactor can be heated by circulation of nitrogen through a 250 kW heater below the core. The nitrogen flows

  15. Damage Modeling Of Injection-Molded Short- And Long-Fiber Thermoplastics

    International Nuclear Information System (INIS)

    Nguyen, Ba Nghiep; Kunc, Vlastimil; Bapanapalli, Satish K.; Phelps, Jay; Tucker, Charles L. III

    2009-01-01

    This article applies the recent anisotropic rotary diffusion - reduced strain closure (ARD-RSC) model for predicting fiber orientation and a new damage model for injection-molded long-fiber thermoplastics (LFTs) to analyze progressive damage leading to total failure of injection-molded long-glass-fiber/polypropylene (PP) specimens. The ARD-RSC model was implemented in a research version of the Autodesk Moldflow Plastics Insight (MPI) processing code, and it has been used to simulate injection-molding of a long-glass-fiber/PP plaque. The damage model combines micromechanical modeling with a continuum damage mechanics description to predict the nonlinear behavior due to plasticity coupled with damage in LFTs. This model has been implemented in the ABAQUS finite element code via user-subroutines and has been used in the damage analyses of tensile specimens removed from the injection-molded long-glass-fiber/PP plaques. Experimental characterization and mechanical testing were performed to provide input data to support and validate both process modeling and damage analyses. The predictions are in agreement with the experimental results.

  16. Oxidative damage of DNA in subjects occupationally exposed to lead.

    Science.gov (United States)

    Pawlas, Natalia; Olewińska, Elżbieta; Markiewicz-Górka, Iwona; Kozłowska, Agnieszka; Januszewska, Lidia; Lundh, Thomas; Januszewska, Ewa; Pawlas, Krystyna

    2017-09-01

    Exposure to lead (Pb) in environmental and occupational settings continues to be a serious public health problem and may pose an elevated risk of genetic damage. The aim of this study was to assess the level of oxidative stress and DNA damage in subjects occupationally exposed to lead. We studied a population of 78 male workers exposed to lead in a lead and zinc smelter and battery recycling plant and 38 men from a control group. Blood lead levels were detected by graphite furnace atomic absorption spectrophotometry and plasma lead levels by inductively coupled plasma-mass spectrometry. The following assays were performed to assess the DNA damage and oxidative stress: comet assay, determination of 8-hydroxy-2'-deoxyguanosine (8-OHdG), lipid peroxidation and total antioxidant status (TAS). The mean concentration of lead in the blood of the exposed group was 392 ± 103 μg/L and was significantly higher than in the control group (30.3 ± 29.4 μg/L, p lead exposure [lead in blood, lead in plasma, zinc protoporphyrin (ZPP)] and urine concentration of 8-OHdG. The level of oxidative damage of DNA was positively correlated with the level of lipid peroxidation (TBARS) and negatively with total anti-oxidative status (TAS). Our study suggests that occupational exposure causes an increase in oxidative damage to DNA, even in subjects with relatively short length of service (average length of about 10 years). 8-OHdG concentration in the urine proved to be a sensitive and non-invasive marker of lead induced genotoxic damage.

  17. Transport of nuclides during a core meltdown accident, with consideration of filtered venting

    International Nuclear Information System (INIS)

    Haeggblom, H.

    1981-01-01

    A BWR core meltdown accident has been studied with respect to the transport of radioactive and nonactive gases and aerosols. A system consisting of a containment with an outer stone condenser in three parts was considered. Calculations of the aerosol behaviour have been made with the computer programme NAUA and HAARM-3, assuming one single compartment. Results from these calculations have been used for multicompartment calculations with CORRAL II. The code was modified so that particles of different sizes could be considered in the different compartments, and the time dependence of the particles can be arbitrary. In addition to the aerosol transport and deposition, the corresponding quantities for elemental iodine were calculated. It was concluded, that if the total volume of the condenser system is of the order of 10 5 m 3 , practically all elemental iodine and particles can be retained in the system. The only leakage to the environment will be caused by inefficient sealing during the first five hours. The pressure can never damage the condenser. (author)

  18. Earth's inner core: Innermost inner core or hemispherical variations?

    NARCIS (Netherlands)

    Lythgoe, K. H.; Deuss, A.|info:eu-repo/dai/nl/412396610; Rudge, J. F.; Neufeld, J. A.

    2014-01-01

    The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

  19. Dependence of Core and Extended Flux on Core Dominance ...

    Indian Academy of Sciences (India)

    Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

  20. Moving away from exhaustion: how core self-evaluations influence academic burnout.

    Science.gov (United States)

    Lian, Penghu; Sun, Yunfeng; Ji, Zhigang; Li, Hanzhong; Peng, Jiaxi

    2014-01-01

    Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction.

  1. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. On-line core monitoring with CORE MASTER / PRESTO

    International Nuclear Information System (INIS)

    Lindahl, S.O.; Borresen, S.; Ovrum, S.

    1986-01-01

    Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

  3. Building of the system for managing and analyzing the hyperspectral data of drilling core

    International Nuclear Information System (INIS)

    Huang Yanju; Zhang Jielin; Wang Junhu

    2010-01-01

    Drilling core logging is very important for geological exploration, hyperspectral detection provides a totally new method for drilling core logging. To use and analyze the drilling core data more easily, and especially store them permanently, a system is built for analyzing and managing the hyperspectral data. The system provides a convenient way to sort the core data, and extract the spectral characteristics, which is the basis for the following mineral identification. (authors)

  4. Putative photoacoustic damage in skin induced by pulsed ArF excimer laser

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, S.; Flotte, T.J.; McAuliffe, D.J.; Jacques, S.L.

    1988-05-01

    Argon-fluoride excimer laser ablation of guinea pig stratum corneum causes deeper tissue damage than expected for thermal or photochemical mechanisms, suggesting that photoacoustic waves have a role in tissue damage. Laser irradiation (193 nm, 14-ns pulse) at two different radiant exposures, 62 and 156 mJ/cm2 per pulse, was used to ablate the 15-microns-thick stratum corneum of the skin. Light and electron microscopy of immediate biopsies demonstrated damage to fibroblasts as deep as 88 and 220 microns, respectively, below the ablation site. These depths are far in excess of the optical penetration depth of 193-nm light (1/e depth = 1.5 micron). The damage is unlikely to be due to a photochemical mechanism because (a) the photons will not penetrate to these depths, (b) it is a long distance for toxic photoproducts to diffuse, and (c) damage is proportional to laser pulse intensity and not the total dose that accumulates in the residual tissue; therefore, reciprocity does not hold. Damage due to a thermal mechanism is not expected because there is not sufficient energy deposited in the tissue to cause significant heating at such depths. The damage is most likely due to a photoacoustic mechanism because (a) photoacoustic waves can propagate deep into tissue, (b) the depth of damage increases with increasing laser pulse intensity rather than with increasing total residual energy, and (c) the effects are immediate. These effects should be considered in the evaluation of short pulse, high peak power laser-tissue interactions.

  5. Single-mode 37-core fiber with a cladding diameter of 248 μm

    DEFF Research Database (Denmark)

    Sasaki, Y.; Takenaga, K.; Aikawa, K.

    2017-01-01

    A heterogeneous single-mode 37-core fiber with a cladding diameter of 248 μm is designed and fabricated. The fiber provides the highest core count and low total-crosstalk less than −20 dB/1000 km in C+L band....

  6. Study of damages caused by drill fluids in sandstone samples using energy dispersive X-ray fluorescence (EDXRF); Estudo dos danos causados por fluido de perfuracao em amostras de arenito utilizando a fluorescencia de raios X dispersiva em energia (EDXRF)

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Joao Luiz B.; Lopes, Ricardo T.; Anjos, Marcelino J. dos; Leite, Jaques C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Lab. de Instrumentacao Nuclear; Queiroz Neto, Joao C. [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas

    2002-07-01

    The performance of an X ray fluorescence (EDXRF) system was evaluated for the determination of solids invasion profile. The EDXRF was used to identify the elements, that are present in the rock core after the damage tests. Results are presented of an experimental study of formation permeability damage caused by invasion of the inert and extraneous solids simulated by CaCO{sub 3} particles. Laboratory damage tests were performed using two types of sandstones core. Vila Velha, low permeability (15 mD) and Rio Bonito, high permeability (800 mD). The damage tests were performed with utilization of SIRF-P, which simulates the conditions find in-situ (petroleum well). (author)

  7. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  8. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  9. Ex-core fuel damage event at paks causes, consequences and lessons learned

    International Nuclear Information System (INIS)

    Bajsz, J.; Gado, J.

    2004-01-01

    On April 10, 2003 Paks NPP experienced a loss of decay-heat removal to 30 irradiated fuel assemblies undergoing a cleaning process in a fuel service pit near the unit 2 spent fuel pool. Following chemical cleaning of high decay-heat fuel, a delay in removing the cleaning vessel's lid left the cleaning system in such a condition that did not provide adequate cooling to the fuel. After several hours of the fuel being under-cooled, a steam bubble developed in the vessel, essentially uncovering the fuel. When the lid of the vessel was removed, the sudden introduction of cool water thermally shocked the fuel causing significant structural damage and a release of fission product gases to the reactor building. The paper will discuss the causes of the event as well as the contributing factors to it. Detailed information will be given about the planning and preparation of the recovery actions. The in-depth analyses of the consequences and lessons learned complete the lecture. (author)

  10. Methodology development for estimating support behavior of spacer grid spring in core

    International Nuclear Information System (INIS)

    Yoon, Kyung Ho; Kang, Heung Seok; Kim, Hyung Kyu; Song, Kee Nam

    1998-04-01

    The fuel rod (FR) support behavior is changed during operation resulting from effects such as clad creep-down, spring force relaxation due to irradiation, and irradiation growth of spacer straps in accordance with time or increase of burnup. The FR support behavior is closely associated with time or increase of burnup. The FR support behavior is closely associated with FR damage due to fretting, therefore the analysis on the FR support behavior is normally required to minimize the damage. The characteristics of the parameters, which affect the FR support behavior, and the methodology developed for estimating the FR support behavior in the reactor core are described in this work. The FR support condition for the KOFA (KOrean Fuel Assembly) fuel has been analyzed by this method, and the results of the analysis show that the fuel failure due to the fuel rod fretting wear is closely related to the support behavior of FR in the core. Therefore, the present methodology for estimating the FR support condition seems to be useful for estimating the actual FR support condition. In addition, the optimization seems to be a reliable tool for establishing the optimal support condition on the basis of these results. (author). 15 refs., 3 tabs., 26 figs

  11. Core vocabulary of young children with Down syndrome.

    Science.gov (United States)

    Deckers, Stijn R J M; Van Zaalen, Yvonne; Van Balkom, Hans; Verhoeven, Ludo

    2017-06-01

    The aim of this study was to develop a core vocabulary list for young children with intellectual disabilities between 2 and 7 years of age because data from this population are lacking in core vocabulary literature. Children with Down syndrome are considered one of the most valid reference groups for researching developmental patterns in children with intellectual disabilities; therefore, spontaneous language samples of 30 Dutch children with Down syndrome were collected during three different activities with multiple communication partners (free play with parents, lunch- or snack-time at home or at school, and speech therapy sessions). Of these children, 19 used multimodal communication, primarily manual signs and speech. Functional word use in both modalities was transcribed. The 50 most frequently used core words accounted for 67.2% of total word use; 16 words comprised core vocabulary, based on commonality. These data are consistent with similar studies related to the core vocabularies of preschoolers and toddlers with typical development, although the number of nouns present on the core vocabulary list was higher for the children in the present study. This finding can be explained by manual sign use of the children with Down syndrome and is reflective of their expressive vocabulary ages.

  12. The biospeckle method for early damage detection of fruits

    Science.gov (United States)

    Yan, Lei; Liu, Jiaxin; Men, Sen

    2017-07-01

    In the field of fruits damage assessment, biospeckle activity is considered relevant to quality properties of plants, such us damage, aging, or diseases. In this paper, biospeckle technique was applied to identify the early bruising of apples. Then a total of 50 undamaged apples were determined to be artificially bruised as samples. Three methods (Fujii, GD, and LSTCA) were used to extract effective information from these speckle images for measuring the intensity of biospeckle activity. The results showed that for all of three methods, the biospeckle activities of the undamaged areas in apple were similar; after the hit, the damaged area showed a lower biospeckle activity. It can be concluded that early bruising can be identified by biospeckle technique.

  13. Permeability of WIPP Salt During Damage Evolution and Healing

    International Nuclear Information System (INIS)

    BODNER, SOL R.; CHAN, KWAI S.; MUNSON, DARRELL E.

    1999-01-01

    The presence of damage in the form of microcracks can increase the permeability of salt. In this paper, an analytical formulation of the permeability of damaged rock salt is presented for both initially intact and porous conditions. The analysis shows that permeability is related to the connected (i.e., gas accessible) volumetric strain and porosity according to two different power-laws, which may be summed to give the overall behavior of a porous salt with damage. This relationship was incorporated into a constitutive model, known as the Multimechanism Deformation Coupled Fracture (MDCF) model, which has been formulated to describe the inelastic flow behavior of rock salt due to coupled creep, damage, and healing. The extended model was used to calculate the permeability of rock salt from the Waste Isolation Pilot Plant (WIPP) site under conditions where damage evolved with stress over a time period. Permeability changes resulting from both damage development under deviatoric stresses and damage healing under hydrostatic pressures were considered. The calculated results were compared against experimental data from the literature, which indicated that permeability in damaged intact WIPP salt depends on the magnitude of the gas accessible volumetric strain and not on the total volumetric strain. Consequently, the permeability of WIPP salt is significantly affected by the kinetics of crack closure, but shows little dependence on the kinetics of crack removal by sintering

  14. Enhanced Core Noise Modeling for Turbofan Engines

    Science.gov (United States)

    Stone, James R.; Krejsa, Eugene A.; Clark, Bruce J.

    2011-01-01

    This report describes work performed by MTC Technologies (MTCT) for NASA Glenn Research Center (GRC) under Contract NAS3-00178, Task Order No. 15. MTCT previously developed a first-generation empirical model that correlates the core/combustion noise of four GE engines, the CF6, CF34, CFM56, and GE90 for General Electric (GE) under Contract No. 200-1X-14W53048, in support of GRC Contract NAS3-01135. MTCT has demonstrated in earlier noise modeling efforts that the improvement of predictive modeling is greatly enhanced by an iterative approach, so in support of NASA's Quiet Aircraft Technology Project, GRC sponsored this effort to improve the model. Since the noise data available for correlation are total engine noise spectra, it is total engine noise that must be predicted. Since the scope of this effort was not sufficient to explore fan and turbine noise, the most meaningful comparisons must be restricted to frequencies below the blade passage frequency. Below the blade passage frequency and at relatively high power settings jet noise is expected to be the dominant source, and comparisons are shown that demonstrate the accuracy of the jet noise model recently developed by MTCT for NASA under Contract NAS3-00178, Task Order No. 10. At lower power settings the core noise became most apparent, and these data corrected for the contribution of jet noise were then used to establish the characteristics of core noise. There is clearly more than one spectral range where core noise is evident, so the spectral approach developed by von Glahn and Krejsa in 1982 wherein four spectral regions overlap, was used in the GE effort. Further analysis indicates that the two higher frequency components, which are often somewhat masked by turbomachinery noise, can be treated as one component, and it is on that basis that the current model is formulated. The frequency scaling relationships are improved and are now based on combustor and core nozzle geometries. In conjunction with the Task

  15. Moving Away from Exhaustion: How Core Self-Evaluations Influence Academic Burnout

    Science.gov (United States)

    Lian, Penghu; Sun, Yunfeng; Ji, Zhigang; Li, Hanzhong; Peng, Jiaxi

    2014-01-01

    Background Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. Objective The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. Methods A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. Results Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. Conclusions Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction. PMID:24489857

  16. Moving away from exhaustion: how core self-evaluations influence academic burnout.

    Directory of Open Access Journals (Sweden)

    Penghu Lian

    Full Text Available BACKGROUND: Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. OBJECTIVE: The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. METHODS: A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. RESULTS: Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. CONCLUSIONS: Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction.

  17. Study of the structural damage in a niobium-microalloyed steel sheet

    International Nuclear Information System (INIS)

    Fernandes, J.; Riba, J.; Verdeja, J.I.

    1986-01-01

    A quantitative experimental study of the damage developed as a consequence of straining has been performed on a microalloyed (niobium) steel sheet by means of a SEM. Equivalent strains range between 0 and 0.68 and strain paths between 0 and 1 and have been obtained in a bulge test. Damage associated to Al 2 O 3 and SMn inclusions is already present in the ''as received'' sheet and grows with strain. Damage associated to CFe 3 second phase particles appears later in the forming of the sheet. For stages previous to necking SMn stringers have dramatically developed more than 50% of total damage. The nucleation equivalent strain is between 0,3 and 0,4. (author)

  18. Analysis of Damage in Laminated Architectural Glazing Subjected to Wind Loading and Windborne Debris Impact

    Directory of Open Access Journals (Sweden)

    Daniel S. Stutts

    2013-05-01

    Full Text Available Wind loading and windborne debris (missile impact are the two primary mechanisms that result in window glazing damage during hurricanes. Wind-borne debris is categorized into two types: small hard missiles; such as roof gravel; and large soft missiles representing lumber from wood-framed buildings. Laminated architectural glazing (LAG may be used in buildings where impact resistance is needed. The glass plies in LAG undergo internal damage before total failure. The bulk of the published work on this topic either deals with the stress and dynamic analyses of undamaged LAG or the total failure of LAG. The pre-failure damage response of LAG due to the combination of wind loading and windborne debris impact is studied. A continuum damage mechanics (CDM based constitutive model is developed and implemented via an axisymmetric finite element code to study the failure and damage behavior of laminated architectural glazing subjected to combined loading of wind and windborne debris impact. The effect of geometric and material properties on the damage pattern is studied parametrically.

  19. Preliminary Estimation of Local Bypass Flow Gap Sizes for a Prismatic VHTR Core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Jo, Chang Keun; Lee, Won Jae

    2009-01-01

    The Very High Temperature Reactor (VHTR) has been selected for the Nuclear Hydrogen Development and Demonstration (NHDD) project. In the VHTR design, core bypass flow has been one of key issues for core thermal margins and target temperature of the core outlet. The core bypass flow in the prismatic VHTR varies with the core life due to the irradiation shrinkage/ swelling and thermal expansion of the graphite blocks, which could be a significant proportion of the total core flow. Thus, accurate prediction of the bypass flow is of major importance in assuring the core thermal margin. To predict the bypass flow, first of all, local gap sizes between graphite blocks in the core should be determined. The objectives of this work are to develop a methodology for determining the gap sizes and to perform a preliminary evaluation for a reference reactor

  20. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  1. Ultrafast Mid-IR Nonlinear Optics in Gas-filled Hollow-core Photonic Crystal Fibers

    DEFF Research Database (Denmark)

    Habib, Selim

    Invention of hollow-core fiber has been proven an ideal medium to study light-gas interaction. Tight confinement of light inside hollowcore fiber allows unremitting and tailored interaction between light and gas over long distances. In this work, we used a special kind of hollowcore fiber − hollow......-core anti-resonant (HC-AR) fiber to study the various nonlinear effects filled with Raman free noble gas. One of the main striking features of HC-AR fiber is that ∼99.99% light can be guided inside the central hollow-core region, which significantly enhances damage threshold level. HC-AR fiber can sustain...... be tuned by simply changing the pressure of the gas while at the same time providing extremely wide transparency ranges. In this thesis, we propose several low-loss broadband guidance HC-AR fibers and investigate soliton-plasma dynamics using HC-AR fiber filled with noble gas in the mid-IR. The combined...

  2. Total Dose Effects on Bipolar Integrated Circuits at Low Temperature

    Science.gov (United States)

    Johnston, A. H.; Swimm, R. T.; Thorbourn, D. O.

    2012-01-01

    Total dose damage in bipolar integrated circuits is investigated at low temperature, along with the temperature dependence of the electrical parameters of internal transistors. Bandgap narrowing causes the gain of npn transistors to decrease far more at low temperature compared to pnp transistors, due to the large difference in emitter doping concentration. When irradiations are done at temperatures of -140 deg C, no damage occurs until devices are warmed to temperatures above -50 deg C. After warm-up, subsequent cooling shows that damage is then present at low temperature. This can be explained by the very strong temperature dependence of dispersive transport in the continuous-time-random-walk model for hole transport. For linear integrated circuits, low temperature operation is affected by the strong temperature dependence of npn transistors along with the higher sensitivity of lateral and substrate pnp transistors to radiation damage.

  3. Test simulation of neutron damage to electronic components using accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    King, D.B., E-mail: dbking@sandia.gov; Fleming, R.M.; Bielejec, E.S.; McDonald, J.K.; Vizkelethy, G.

    2015-12-15

    The purpose of this work is to demonstrate equivalent bipolar transistor damage response to neutrons and silicon ions. We report on irradiation tests performed at the White Sands Missile Range Fast Burst Reactor, the Sandia National Laboratories (SNL) Annular Core Research Reactor, the SNL SPHINX accelerator, and the SNL Ion Beam Laboratory using commercial silicon npn bipolar junction transistors (BJTs) and III–V Npn heterojunction bipolar transistors (HBTs). Late time and early time gain metrics as well as defect spectra measurements are reported.

  4. Rosiglitazone attenuates pulmonary fibrosis and radiation-induced intestinal damage

    International Nuclear Information System (INIS)

    Mangoni, M.; Gerini, C.; Sottili, M.; Cassani, S.; Stefania, G.; Biti, G.; Castiglione, F.; Vanzi, E.; Bottoncetti, A.; Pupi, A.

    2011-01-01

    Full text of publication follows: Purpose.-The aim of the study was to evaluate radioprotective effect of rosiglitazone (RGZ) on a murine model of late pulmonary damage and of acute intestinal damage. Methods.- Lung fibrosis: C57 mice were treated with the radiomimetic agent bleomycin, with or without rosiglitazone (5 mg/kg/day). To obtain an independent qualitative and quantitative measure for lung fibrosis we used high resolution CT, performed twice a week during the entire observation period. Hounsfield Units (HU) of section slides from the upper and lower lung region were determined. On day 31 lungs were collected for histological analysis. Acute intestinal damage: mice underwent 12 Gy total body irradiation with or without rosiglitazone. Mice were sacrificed 24 or 72 h after total body irradiation and ileum and colon were collected. Results.- Lung fibrosis: after bleomycin treatment, mice showed typical CT features of lung fibrosis, including irregular septal thickening and patchy peripheral reticular abnormalities. Accordingly, HU lung density was dramatically increased. Rosiglitazone markedly attenuated the radiological signs of fibrosis and strongly inhibited HU lung density increase (60% inhibition at the end of the observation period). Histological analysis revealed that in bleomycin-treated mice, fibrosis involved 50-55% of pulmonary parenchyma and caused an alteration of the alveolar structures in 10% of parenchyma, while in rosiglitazone-treated mice, fibrosis involved only 20-25% of pulmonary parenchyma, without alterations of the alveolar structures. Acute intestinal damage: 24 h after 12 Gy of total body irradiation intestinal mucosa showed villi shortening, mucosal thickness and crypt necrotic changes. Rosiglitazone showed a histological improvement of tissue structure, with villi and crypts normalization and oedema reduction. Conclusion.- These results demonstrate that rosiglitazone displays a protective effect on pulmonary fibrosis and radiation

  5. Rosiglitazone attenuates pulmonary fibrosis and radiation-induced intestinal damage

    Energy Technology Data Exchange (ETDEWEB)

    Mangoni, M.; Gerini, C.; Sottili, M.; Cassani, S.; Stefania, G.; Biti, G. [Radiotherapy Unit, Clinical Physiopathology Department, University of Florence, Firenze (Italy); Castiglione, F. [Department of Human Pathology and Oncology, University of Florence, Firenze (Italy); Vanzi, E.; Bottoncetti, A.; Pupi, A. [Nuclear Medicine Unit, Clinical Physiopathology Department, University of Florence, Firenze (Italy)

    2011-10-15

    Full text of publication follows: Purpose.-The aim of the study was to evaluate radioprotective effect of rosiglitazone (RGZ) on a murine model of late pulmonary damage and of acute intestinal damage. Methods.- Lung fibrosis: C57 mice were treated with the radiomimetic agent bleomycin, with or without rosiglitazone (5 mg/kg/day). To obtain an independent qualitative and quantitative measure for lung fibrosis we used high resolution CT, performed twice a week during the entire observation period. Hounsfield Units (HU) of section slides from the upper and lower lung region were determined. On day 31 lungs were collected for histological analysis. Acute intestinal damage: mice underwent 12 Gy total body irradiation with or without rosiglitazone. Mice were sacrificed 24 or 72 h after total body irradiation and ileum and colon were collected. Results.- Lung fibrosis: after bleomycin treatment, mice showed typical CT features of lung fibrosis, including irregular septal thickening and patchy peripheral reticular abnormalities. Accordingly, HU lung density was dramatically increased. Rosiglitazone markedly attenuated the radiological signs of fibrosis and strongly inhibited HU lung density increase (60% inhibition at the end of the observation period). Histological analysis revealed that in bleomycin-treated mice, fibrosis involved 50-55% of pulmonary parenchyma and caused an alteration of the alveolar structures in 10% of parenchyma, while in rosiglitazone-treated mice, fibrosis involved only 20-25% of pulmonary parenchyma, without alterations of the alveolar structures. Acute intestinal damage: 24 h after 12 Gy of total body irradiation intestinal mucosa showed villi shortening, mucosal thickness and crypt necrotic changes. Rosiglitazone showed a histological improvement of tissue structure, with villi and crypts normalization and oedema reduction. Conclusion.- These results demonstrate that rosiglitazone displays a protective effect on pulmonary fibrosis and radiation

  6. Damage detection in high-rise buildings using damage-induced rotations

    International Nuclear Information System (INIS)

    Sung, Seung Hun; Jung, Ho Youn; Lee, Jung Hoon; Jung, Hyung Jo

    2016-01-01

    In this paper, a new damage-detection method based on structural vibration is proposed. The essence of the proposed method is the detection of abrupt changes in rotation. Damage-induced rotation (DIR), which is determined from the modal flexibility of the structure, initially occurs only at a specific damaged location. Therefore, damage can be localized by evaluating abrupt changes in rotation. We conducted numerical simulations of two damage scenarios using a 10-story cantilever-type building model. Measurement noise was also considered in the simulation. We compared the sensitivity of the proposed method to localize damage to that of two conventional modal-flexibility-based damage-detection methods, i.e., uniform load surface (ULS) and ULS curvature. The proposed method was able to localize damage in both damage scenarios for cantilever structures, but the conventional methods could not

  7. Damage detection in high-rise buildings using damage-induced rotations

    International Nuclear Information System (INIS)

    Sung, Seung Hoon; Jung, Ho Youn; Lee, Jung Hoon; Jung, Hyung Jo

    2014-01-01

    In this paper, a new damage-detection method based on structural vibration is proposed. The essence of the proposed method is the detection of abrupt changes in rotation. Damage-induced rotation (DIR), which is determined from the modal flexibility of the structure, initially occurs only at a specific damaged location. Therefore, damage can be localized by evaluating abrupt changes in rotation. We conducted numerical simulations of two damage scenarios using a 10-story cantilever-type building model. Measurement noise was also considered in the simulation. We compared the sensitivity of the proposed method to localize damage to that of two conventional modal-flexibility-based damage-detection methods, i.e., uniform load surface (ULS) and ULS curvature. The proposed method was able to localize damage in both damage scenarios for cantilever structures, but the conventional methods could not.

  8. Dynamico-FE: A Structure-Preserving Hydrostatic Dynamical Core

    Science.gov (United States)

    Eldred, Christopher; Dubos, Thomas; Kritsikis, Evaggelos

    2017-04-01

    It is well known that the inviscid, adiabatic equations of atmospheric motion constitute a non-canonical Hamiltonian system, and therefore posses many important conserved quantities such as as mass, potential vorticity and total energy. In addition, there are also key mimetic properties (such as curl grad = 0) of the underlying continuous vector calculus. Ideally, a dynamical core should have similar properties. A general approach to deriving such structure-preserving numerical schemes has been developed under the frameworks of Hamiltonian methods and mimetic discretizations, and over the past decade, there has been a great deal of work on the development of atmospheric dynamical cores using these techniques. An important example is Dynamico, which conserves mass, potential vorticity and total energy; and possesses additional mimetic properties such as a curl-free pressure gradient. Unfortunately, the underlying finite-difference discretization scheme used in Dynamico has been shown to be inconsistent on general grids. To resolve these accuracy issues, a scheme based on mimetic Galerkin discretizations has been developed that achieves higher-order accuracy while retaining the structure-preserving properties of the existing discretization. This presentation will discuss the new dynamical core, termed Dynamico-FE, and show results from a standard set of test cases on both the plane and the sphere.

  9. Pipeline welding with Flux Cored and Metal Cored Wire; Soldagem de dutos com processos Arame Tubular e de Alma Metalica

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Ubirajara Pereira da [ITW Soldagem Brasil Miller-Hobart, Sao Paulo, SP (Brazil)

    2003-07-01

    Different welding process like SMAW, Semi-Automatic FCAW Gas-shielded and Self-shielded and Mechanized GMAW-MAG with Solid Wire are suggested to weld Transmission Pipelines. Presently, the largest extensions of Transmission Pipelines under construction, are in China like Lines West-East, Zong-Wu, Shan-Jing Fuxian and some others, totalizing about 8.000 km, and all using Semi-Automatic Self Shielded Flux Cored Arc Welding Process. Also, several papers and magazines that covers Transmission Pipelines Welding, not frequently mention Operational aspects of the process and some other variables like environment and site geography. This presentation intends to cover some of the Operational aspects of the Flux Cored Arc Welding and GMAW-Metal Cored in order to give sufficient information for Construction, Engineering, Projects e Contractors so they can evaluate these Process against the SMAW or even Mechanized Systems, considering the Operation Factor, Efficiency and Deposition Rate. We will not cover operational details of the GMAW Mechanized Systems but only suggest that be evaluated the possibility to replace the GMAW-Solid Wire by the GMAW-Metal Cored Wire. (author)

  10. DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

    International Nuclear Information System (INIS)

    Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

    1979-01-01

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

  11. Measurement of damage in systemic vasculitis: a comparison of the Vasculitis Damage Index with the Combined Damage Assessment Index

    DEFF Research Database (Denmark)

    Suppiah, Ravi; Flossman, Oliver; Mukhtyar, Chetan

    2011-01-01

    To compare the Vasculitis Damage Index (VDI) with the Combined Damage Assessment Index (CDA) as measures of damage from vasculitis.......To compare the Vasculitis Damage Index (VDI) with the Combined Damage Assessment Index (CDA) as measures of damage from vasculitis....

  12. Thermal margin model for transition core of KSNP

    International Nuclear Information System (INIS)

    Nahm, Kee Yil; Lim, Jong Seon; Park, Sung Kew; Chun, Chong Kuk; Hwang, Sun Tack

    2004-01-01

    The PLUS7 fuel was developed with mixing vane grids for KSNP. For the transition core partly loaded with the PLUS7 fuels, the procedure to set up the optimum thermal margin model of the transition core was suggested by introducing AOPM concept into the screening method which determines the limiting assembly. According to the procedure, the optimum thermal margin model of the first transition core was set up by using a part of nuclear data for the first transition and the homogeneous core with PLUS7 fuels. The generic thermal margin model of PLUS7 fuel was generated with the AOPM of 138%. The overpower penalties on the first transition core were calculated to be 1.0 and 0.98 on the limiting assembly and the generic thermal margin model, respectively. It is not usual case to impose the overpower penalty on reload cores. It is considered that the lack of channel flow due to the difference of pressure drop between PLUS7 and STD fuels results in the decrease of DNBR. The AOPM of the first transition core is evaluated to be about 135% by using the optimum generic thermal margin model which involves the generic thermal margin model and the total overpower penalty. The STD fuel is not included among limiting assembly candidates in the second transition core, because they have much lower pin power than PLUS7 fuels. The reduced number of STD fuels near the limiting assembly candidates the flow from the limiting assembly to increase the thermal margin for the second transition core. It is expected that cycle specific overpower penalties increase the thermal margin for the transition core. Using the procedure to set up the optimum thermal margin model makes sure that the enhanced thermal margin of PLUS7 fuel can be sufficiently applied to not only the homogeneous core but also the transition core

  13. Extent of moisture and mould damage in structures of public buildings

    Directory of Open Access Journals (Sweden)

    Petri J. Annila

    2017-06-01

    Full Text Available The study concentrated on the extent of moisture and mould damage in different structures in 25 public buildings in Finland. Users of all the buildings had health symptoms suspected to be the result of moisture and mould damage, which is why moisture performance assessments had been performed. The assessment reports on each building were available as research material. The reports indicated that the examined buildings suffered from multiple moisture and mould problems in several different structures. On average, however, a relatively small proportion of the total number of structures had suffered damage. On the basis of the research material, damage was most extensive in walls in soil contact (16.3% and base floor structures (12.5%. The lowest damage rates were found in partition walls (2.4%, external walls (2.6% and intermediate floors (2.5%. The results of the study underline the importance of thorough moisture performance assessments to ensure that all point-sized moisture and mould damage is detected.

  14. Metric to quantify white matter damage on brain magnetic resonance images

    Energy Technology Data Exchange (ETDEWEB)

    Valdes Hernandez, Maria del C.; Munoz Maniega, Susana; Anblagan, Devasuda; Bastin, Mark E.; Wardlaw, Joanna M. [University of Edinburgh, Department of Neuroimaging Sciences, Centre for Clinical Brain Sciences, Edinburgh (United Kingdom); University of Edinburgh, Centre for Cognitive Ageing and Cognitive Epidemiology, Edinburgh (United Kingdom); UK Dementia Research Institute, Edinburgh Dementia Research Centre, London (United Kingdom); Chappell, Francesca M.; Morris, Zoe; Sakka, Eleni [University of Edinburgh, Department of Neuroimaging Sciences, Centre for Clinical Brain Sciences, Edinburgh (United Kingdom); UK Dementia Research Institute, Edinburgh Dementia Research Centre, London (United Kingdom); Dickie, David Alexander; Royle, Natalie A. [University of Edinburgh, Department of Neuroimaging Sciences, Centre for Clinical Brain Sciences, Edinburgh (United Kingdom); University of Edinburgh, Centre for Cognitive Ageing and Cognitive Epidemiology, Edinburgh (United Kingdom); Armitage, Paul A. [University of Sheffield, Department of Cardiovascular Sciences, Sheffield (United Kingdom); Deary, Ian J. [University of Edinburgh, Centre for Cognitive Ageing and Cognitive Epidemiology, Edinburgh (United Kingdom); University of Edinburgh, Department of Psychology, Edinburgh (United Kingdom)

    2017-10-15

    Quantitative assessment of white matter hyperintensities (WMH) on structural Magnetic Resonance Imaging (MRI) is challenging. It is important to harmonise results from different software tools considering not only the volume but also the signal intensity. Here we propose and evaluate a metric of white matter (WM) damage that addresses this need. We obtained WMH and normal-appearing white matter (NAWM) volumes from brain structural MRI from community dwelling older individuals and stroke patients enrolled in three different studies, using two automatic methods followed by manual editing by two to four observers blind to each other. We calculated the average intensity values on brain structural fluid-attenuation inversion recovery (FLAIR) MRI for the NAWM and WMH. The white matter damage metric is calculated as the proportion of WMH in brain tissue weighted by the relative image contrast of the WMH-to-NAWM. The new metric was evaluated using tissue microstructure parameters and visual ratings of small vessel disease burden and WMH: Fazekas score for WMH burden and Prins scale for WMH change. The correlation between the WM damage metric and the visual rating scores (Spearman ρ > =0.74, p < 0.0001) was slightly stronger than between the latter and WMH volumes (Spearman ρ > =0.72, p < 0.0001). The repeatability of the WM damage metric was better than WM volume (average median difference between measurements 3.26% (IQR 2.76%) and 5.88% (IQR 5.32%) respectively). The follow-up WM damage was highly related to total Prins score even when adjusted for baseline WM damage (ANCOVA, p < 0.0001), which was not always the case for WMH volume, as total Prins was highly associated with the change in the intense WMH volume (p = 0.0079, increase of 4.42 ml per unit change in total Prins, 95%CI [1.17 7.67]), but not with the change in less-intense, subtle WMH, which determined the volumetric change. The new metric is practical and simple to calculate. It is robust to variations in

  15. Estimation and Control for Autonomous Coring from a Rover Manipulator

    Science.gov (United States)

    Hudson, Nicolas; Backes, Paul; DiCicco, Matt; Bajracharya, Max

    2010-01-01

    A system consisting of a set of estimators and autonomous behaviors has been developed which allows robust coring from a low-mass rover platform, while accommodating for moderate rover slip. A redundant set of sensors, including a force-torque sensor, visual odometry, and accelerometers are used to monitor discrete critical and operational modes, as well as to estimate continuous drill parameters during the coring process. A set of critical failure modes pertinent to shallow coring from a mobile platform is defined, and autonomous behaviors associated with each critical mode are used to maintain nominal coring conditions. Autonomous shallow coring is demonstrated from a low-mass rover using a rotary-percussive coring tool mounted on a 5 degree-of-freedom (DOF) arm. A new architecture of using an arm-stabilized, rotary percussive tool with the robotic arm used to provide the drill z-axis linear feed is validated. Particular attention to hole start using this architecture is addressed. An end-to-end coring sequence is demonstrated, where the rover autonomously detects and then recovers from a series of slip events that exceeded 9 cm total displacement.

  16. Gamma radiation damage in pixelated detector based on carbon nanotubes

    International Nuclear Information System (INIS)

    Leyva, A.; Pinnera, I.; Leyva, D.; Abreu, Y.; Cruz, C. M.

    2013-01-01

    The aim of this paper is to evaluate the possible gamma radiation damage in high pixelated based on multi-walled carbon nanotubes detectors, grown on two different substrata, when it is operating in aggressive radiational environments. The radiation damage in displacements per atom (dpa) terms were calculated using the MCCM algorithm, which takes into account the McKinley-Feshbach approach with the Kinchin-Pease approximation for the damage function. Was observed that with increasing of the gamma energy the displacement total number grows monotonically reaching values of 0.39 displacements for a 10 MeV incident photon. The profiles of point defects distributions inside the carbon nanotube pixel linearly rise with depth, increasing its slope with photon energy. In the 0.1 MeV - 10 MeV studied energy interval the electron contribution to the total displacement number become higher than the positron ones, reaching this last one a maximum value of 12% for the 10 MeV incident photons. Differences between the calculation results for the two used different substrata were not observed. (Author)

  17. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    CERN Document Server

    Kirchner, G

    1981-01-01

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

  18. Space Radiation Induced Cytogenetic Damage in the Blood Lymphocytes of Astronauts: Persistence of Damage After Flight and the Effects of Repeat Long Duration Missions

    Science.gov (United States)

    George, Kerry; Rhone, Jordan; Chappell, L. J.; Cucinotta, F. A.

    2010-01-01

    Cytogenetic damage was assessed in blood lymphocytes from astronauts before and after they participated in long-duration space missions of three months or more. The frequency of chromosome damage was measured by fluorescence in situ hybridization (FISH) chromosome painting before flight and at various intervals from a few days to many months after return from the mission. For all individuals, the frequency of chromosome exchanges measured within a month of return from space was higher than their prefight yield. However, some individuals showed a temporal decline in chromosome damage with time after flight. Statistical analysis using combined data for all astronauts indicated a significant overall decreasing trend in total chromosome exchanges with time after flight, although this trend was not seen for all astronauts and the yield of chromosome damage in some individuals actually increased with time after flight. The decreasing trend in total exchanges was slightly more significant when statistical analysis was restricted to data collected more than 220 days after return from flight. In addition, limited data on multiple flights show a lack of correlation between time in space and translocation yields. Data from three crewmembers who has participated in two separate long-duration space missions provide limited information on the effect of repeat flights and show a possible adaptive response to space radiation exposure.

  19. Glaciological and chemical studies on ice cores from Hans Tausen ice cap, Greenland

    DEFF Research Database (Denmark)

    Clausen, H.B.; Stampe, Mia; Hammer, C.U.

    2001-01-01

    The paper presents studies of various chemical and isotopical parameters from ice cores drilled in the northernmost located ice cap, Hans Tausen Iskappe, Pearyland, Greenland (HT). The 346 m main core (MC95) was drilled to bedrock in 1995 as well as a 35 m shallow core (SC95). A 60 m shallow core...... (SC75) and a 51 m shallow core (SC76) was drilled at two different positions in 1975 and 1976, respectively. A 6 m shallow core (SC94) was drilled in 1994. Continuous stable isotope records exist for all of these cores, total b-activity only from SC75 and SC76. Continuous ECM inferred acidity records...... exist along the 1995 cores (MC95 and SC95) and finally detailed records of dust and water soluble ion concentrations exist on selected parts of MC95. To determine a time scale for the ice core is an important prerequisite for the interpretation of other records. The age scale is based on acid layers...

  20. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by ins